NUREG/IA-O 122

ICSP-GA-MSIV-T

International

Agreement Report

Assessment of MSIV Full

Closure for Santa Maria

De Garoila Nuclear Power Plant

Using TRAC-BFl (GiJi)

Prepared by

J. L- Crespo/University of Cantabria

R. A. Femandez/Nuclenor, S. A.

University of Cantabria

FAG Ciencias. Seccion De Fisicas. Nuclenor

c/Hernan Cortes, 26

39080 Santander, 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 theregf, 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/JA-0122

ICSP-GA-MSIV-T

International

Agreement Report

Assessment of MSIV Full

Closure for Santa Maria

De Garon-a Nuclear Power Plant

Using TRAC-BF1 (GiJi)

Prepared by

J. L CrespolUniversity of Cantabria

R. A. Fernanrdez/Nuclenor, S. A.

University of Cantabria

FAC Ciencias. Seccion De Fisicas. Nuclenor

cfHernan Cortes, 26

39080 Santander, 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, orr represents that

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

Page

EXECUTIVE SUMMARY......................................................... V

ABSTRACT...................................................................... Vii

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

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

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

It.- PLANT AND TRANSIENT DESCRIPTION ....................................... 2

Ill.- CODE INPUT MODEL DESCRIPTION......................................... 4

IV.- CALCULATION RESULTS...................................................... 8

V.- RUN STATISTICS............................................................... 10

VI.- CONCLUSIONS ............... ........... 1

VII.- REFERENCES .................................................................. 12

iii

EXECUTIVE SUMMARY

An assessment of the first 60 seconds of a spurious Main Steam Isolation Valve (MSIV's)

closure for Santa Maria de Garofla Nuclear Power Plant using TRAC-BF11 code is presented.

Reasonable and realistic adjustments have been made in the model to improve its perfomance.

This work is part of the validlation set for the .TRAC model that it is being developed for wider

use and allow to test the code capabilities.

As a result of the analysis, It is felt that TRAC-BF1 Is capable of reproducing the plant

behaviour with an acceptable degree of accuracy although better models are clearfy needed, in

addition to further noding work and code improvements. The code took almost 14000 sec.

which makes a 1/230 calculatic n time to real time ratio.

For this transient a mechanisti: separator model is needed. It will also help to cut down running

costs if the vessel noding cou~d have different number of cells at different heigths. Though not

very important for this trans;ient, the critical flow model will allow for realistic RV flow

assumptions.

There are not guidelines available for separator modelling in transients. It has been found

that a detailed noding in thel separator region may be needed to represent steam-water

Interaction.

V

ABSTRACT

This document presents a spurious MSIV closure analysis for Santa Maria de Garofla Nuclear

Power Plan describing the problems found when comparing calculated and real data.

The plant is a General Electric Boiling Water Reactor 3, containment type Mark 1. It is operated

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

The analysis has been performed by 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 GaroFia Power Plant.

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

second of two NUCLENOR contributions to the International Code Applications and

Assessment Program (ICAP).

The code was run in a Cyber 932 with operating system NOSNE, property of NUCLENOR,

S.A.. A programming effort was carried out in order to provide suitable graphics from the output

file.

. Vii
.LIST OF TABLES

Page

TABLE 11-1

TABLE 111-1

TABLE IV-11

TABLE IV..2

TABLE V-1

Garofla NPP Design Characteristics

Description of the TRAC components

Initial Conditions

Chronology of Events

Run Statistics

13

15

16

17

18

viii
LIST OF FIGURES

Page

Fig. Il-i Measured Dome Pressure 19

Fig. 11-2 Measured Downcomer water level 20

Fig. 11-3 Measured Core Flow 21

Fig.' 11-4 Measured Feedwater Flow 22

Fig. Ill-i TRAC-BFI Plant Model 23

Fig. 111-2 Vessel model 24

Fig. IV-1 Steam Dome Pressure 25

Fig. IV-2 Mixing plenum void fraction 26

Fig. IV-3 Core flow 27

Fig. IV-4 Jet pump throat void fraction 28

Fig. IV-5 Downcomer water level 29

Fig. IV-6 Basic model results 30

Fig. IV-7 8x2 vessel results 31

Fig. IV-8 Wx vessel results 32

Fig. V-1 TRAC Time Step 33

i X

F 0ORE WOR D

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 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 techni cal 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:

- iUnidad E16ctrica (UNESA)

- Uni6n Iberoamericana de Tecnologia E16ctrica (UITESA)

- Empresa Nacional del Uranio (ENUSA)

- TECNATOM

- Empresarios Agrupados

- LOFT-ESPA~A

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

Coordinacibn". This committee has approved the distribution

of this document for ICAP purposes.

D-1 /90- MPN V

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-BFI 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 assesment case the analysis of the first 60 seconds a spurious

full MSIV closure. The analysis, described in the present report, was carried out by the Applied

Physics Department of University of Cantabria and the Analysis and Operations Section of

NUCLENOR.

The transient consisted of a full MSIV closure at 100% power which actually took place in the

plant, so it will help to time the model for further analysis.

The model was changed to improve its accuracy, but all the changes were made only if there

was a physical support. The modelling accuracy had to trade off with the simplicity needed to

keep down running costs.

A summary description of the Santa Maria de Garofla 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.

1
11. -PLANT AND TRANSIENT DESCRIPTION

Il PLANT DESCRIPTION

Santa Maria de Garofla Nuclear Power Plant Is a General Electric Boiling Water Reactor 3, with

containment type Mark 1. The plant is operated by NUCLENOR, which is a subsidiary of

Iberduero, S.A. and Electra de Viesgo, S.A., and w .as 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 thei 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 (134C). A complete description

of relevant parameters of the plant is shown in Table Il-i.

A Recirculation System, consisting of two external centrifugal recirculation pumps (variable

speed) and twenty reactor vessel Internal let 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 hat mixed with subcooled feedwater entering the vessel at the steam separator

elevation (9.33 m. above the'downcomer bottom). The' recirculation pumps suction exits the'

vessel at an elevat~ion 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 SRV's.

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
A Feedwater System made up for the inventory coming out of the vessel through main steam

line by two out of three motor-driver pumps. A Feedwater Control System assures the water

level in the downcomer remains between predetermined limi~ts.

12 TRANSIENT DESCRIPTION

A spurious Main Steam Isolation Valve (MSIV's) closure signal occurs at time zero. The core

was operating at 100% power and 89% core flow at the start of the transient. No relevant

manual actions took place the Isolation Condenser System is not activated while Feedwater

and Recirculation System continue in operation during the first 60 seconds.

The MSIV's closed completely in about 3 seconds. The SCRAM took place when the valves

were at 90% position. The steam that was still coming out of the core caused a pressure

increase till the RV's setpoints were reached; two of them opened for 3 seconds and the third

one opened for 5 seconds. Even- after the RV's closure the pressure kept decreasing due to

condensation around the feedwater sparger and the separators.

The vessel level fell down because of void collapse until the feedwater flow recovered it.

The core flow varied because of voids redistribution throughout the vessel, getting larger as

it moved from two-phase to single phase pressure drop. This helped to accelerate the' vessel

depresurization and cooling. The drive flow kept the same because of the flow controller being

In manual mode. Figures 11.1 though 11.4 show the most important variables recorded from the

process computer.The instrumentation precision is:

Dome pressure 0.5%

Core flow ', 3%

Downcomer level 0.5%

Feedwater flow .. 2%.,,

3
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-BFI. The

-BREAK and FILL components are used to impose thermiahydraulic 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. 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 this analysis was based on data taken from

drawings, and specific technical documents. (Ref. 1)

The model contains 35 components with 33 fluid juntions. Table Ill-i includes a listing of all

the TRAC components used in the model. Figure 1ll-1 shows a schematic of the TRAC-BF1

model. The illustrated model represents Garofla for normal operation only. If should be noted

that this is a generic model, not specifically related to this transient.

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

shown in Figure 111-2. This nodalization is made up of ten axial levels, three radial segments and

one azimuthal sector.

The radial segments correspond to the separator, surrounding region and downcomer

region.Since only one average bundle is studied, two rings are actually used below the

separator. Though the dryers skirt has a different inner diameter than the downcomer, no extra

segment was considered. The difference is taken into account by volume and area fractions.

This avoids large number of cells in the vessel model

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

4
axial levels (3 and 4) of different lenght. The remaining axial levels 5 through 10, represent the

steam dome, upper plenum, with the feedwater inlet at axial level 6 and the main steam line

outlet at the axial level 9. The downcomer extends from axial level 2 to axial level 5.

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

used in axial level 8. This allows the vapor to continue upwards into the axial level 9 and the

liquid to drain radially outward into the sorrounding region.

The guide tubes are not modelled and their volume has been substracted from the lower

plenum.

Fuel bundle modelling is accomplished using one CHAN component simulating the~active

region of the 400 bundles.

The axial distribution of the CHAN components consists of 12 axial levels. The bypass flow

path is modelled in the VESSEL component. Only an average rod group is modelled.

Axial power profile was taken from plant data recorded 2 days afterwards for the same core

flow and power. 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) built

into the code. 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 2.

Both recirculation loops have been modelled by six TRAC-BF1 component, neglecting the

small differences between them.

- TEE 20: suction piping conected to the vessel that accounts for IC connection.

- VALVE 21: suction valves.

- PUMP 22: main pumps.

- VALVE 23: discharge valves.

-TEE 24: discharge piping accounting for the LPCI connection.

- JETPUMP 25: jet pumps.

The noding through the recirculation system is almost the minimum that takes into account all

connections. The PUMP component represents only the fluid volume between the inlet and

outlet nozzles of the pump. The Internal friction is accounted for In 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 3 via one

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

path system.

5
The main steam Ines are modelled as follows:

- TEE 50:

-VALVE 51:

-TEE 52:'

- VALVE 53:

- VALVE 54:

- TEE 55:

- VALVE 69;

- VALVE 68:

- BREAK 58:

- BREAK 56:

- BREAK 57:

- BREAK 59:

connection to the VESSEL and RV's.

RV's & SRV's.

connection to SV's.

SV's.

MSIV'S.

connection of the bypass line.

TCV's.

Bypass valve.

Turbine back pressure.

Suppression pool pressure.

Drywell pressure.

Bypass back pressure.

The reactor point kinectics option was turned on in 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 2 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 critical flow model was applied at the RV's, adjusting the area in order to match the flow

measured during the startup tests. Trying the nominal area produced 30% more than the

certified capacity, what was considered too high.

Control systems are modelled for the pressure regulator and trip logics.

Although a feedwater control model was available, it was not considered in the analysis

because the performance of the control valves in the trasient resulted a hard task to model.

In order to run the transient, the MSIV trip was activated. The following boundary conditions

were assumed:

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

mnultiplier factor of 1.0.

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

c.- Heat conduction through the vessel intermals i s' neglected except for the bundle wall.

.- The backpressure at the turbine while the MSIV's were closing was assumed constant.

6
e.- The feedwater flow was input from plant data because the control valves were presumably

locked for part of the transient, producing an abnormal response.

f.- Feedwater heaters inertia was assumed to be large enough as to avoid .significant

temperature changes.

The model has been developed by the analysis from simpler to more complex models,

concerning the upper part of the vessel and the separator location. This Is presented as a

sensitivity analysis.

7
IV.. CALCULATION RESULTS

The calculation of the transient with TRAC-BF1/G1J1 has been developed in two steps. First

of all, a steady state calculation was carried out. The second step consisted of the transient

calculation.

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

1 .OE.04, 1 .OE-05, respectively. The maximum number of the outer iterations (QITMAX) was 8

and the VESSEL calculation was solved by direct inversion.

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-5.

The comparison between TRAC and the plant is limited to the available plant data, that is,

dome pressure, core flow, downcomer level and core plate Ap. It was also recorded the

APRM's signal but since the SCRAM takes place in the first second and the data were taken

every 5 seconds, there is no relevant information for this transient in that plot.

Figure IV-1 shows reactor pressure. The pressure peak is well represented. However, after

the RV closure the code predicts a higher depresurization rate, although reaching similar

minimum 10 seconds in advance. This is related to the condensing eff ect of the water flowing

out of the separators, that is the main issue in this calculation.

The separator model used send the water downwards into the inner zone and the steam

upwards into the upper zone. The separator model sends the water radially into the

surrounding region and the steam upwards into the upper cell.

The separator model used have a good performance at rated conditions. After the SCRAM

there is very little power in the core, but there is full core flow and the feedwater is making it

cooler; as a result the core outlet quality gets close to 0. Figure IV-2 shows calculated mixing

plenum void fraction. Simultaneously, due to the SCRAM a large level change takes place.

These two facts are going to affect significantly the separator performance (carryover,

carryunder, pressure drop). Obviously, it is different sending water under the water level and

sending it upwards into the steam upper zone.

The separator model is ideal. No matter what the conditions are, there will not be water flow

upwards or steam flow outwards, and this has not to occur necessarily at off rated conditions.

8
Since the pressure is dependent on the steam-water mixing process, no great accuracy can

be expected with this separator model (other separators models in TRAC do not work properly).

A different analysis was made with a different cure pressure drop. The calculated pressure

was closer to the data but the core flow was clearly lower.

Figure IV-3 shows core flow. The calculated results are good in general, but around 20 sec.

the code produces lower flow not seen in the data. This Is due to cavitation in the let pump as

can be seen in Figure IVA4 which shows the throat void fraction.

Figure IV.5 shows water level in the downcomer. TRAC results refer to collapsed level. The

calculation is good in the first 40 seconds, but gets too high afterwards, with a fin al diff erence of

about 0.5 m. This is believed to be caused by a lack of better modelling f or the flow inside the

dryer skirt, which is related to the separator and upper cell.

The main changes made, when developing the model, affected noding in the upper part of

the vessel and separator location as well as pressure losses distribution. In the following a brief

description of the process is given.

Figure IV-6 shows the results using a much simpler model obtained collapsing radial

segments 2 and 3 into one and axial levels 7, 8 and 9 into one, with the separator located at

level 6, that is, under level 9.33 m. It can be seen that as soon as this is covered the pressure

rises again.

Figure IV-7 shows the results obtained using again 2 radial segments and collapsing levels 6

and 7 into one and 8 and 9 into another one.

The results show the same tendency. The separator discharge was under 9.83 m.

Figure IV-8 corresponds to a model where there are two radial segments and levels 7 and 8

are collapsed Into one with the separator discharge in it.

9
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. The RV's operation and the level

drop seem to be the main cous~es of reduction.

The run statistics are shown In Table V-i. The real fime/CPU time ratio was 11229.

10
VI. C ONCLUSIONS

A good degree of accuracy is attained for the pressure final value although the transient

evolution Is not very close to the data. Nevertheless, this result is surprisingly good for an ideal

separator model, showing that a radial diScharge at separators heigth causes water-steam

mixing not too far from the actual process.

The accuracy in core flow and water level is not so good. Too fast depresurization will take jet

pumps close to cavitation. Longer condensation will mean larger water inventory too. Anyway

these parameters may also be affected by noding problems such as steam numerical diffusion

into the downcomer or improper flow path representation from inside to outside of the dryer

skirt.

In spite of all this TRAC model shows overall acceptable perfomnance, which no doubt will be

improved when a mechanistic separator is available.

By going through the process of improving the model it was found that the EXTRACT

capability Is not as powerful as needed. It should include the control mode.

11
VII. REFERENCES

1- NUCLENOR, S.A. "Cuademos de c~Iculo. NUCLE-01" Rev. 0, Febrero de 1989

2.- ENUSA. "Fvaluaci6n de seguridad dg la recarga. C. N. Santa Marfa de Garofla Ciclo 15".

12
TABLE I I - 1

GAROFIA NPP DESIGN CARACTHERISTICS

R EACTOR

" Thermal Power (100%)

" Vessel Pressure

" Core Flow

" Steam Flow

" Feedwater Temperature

CORE DIMENSIONS

" Diameter

" Active Lenght

1380 Mw

70.3 Kg/cm2

21.77 x 106 Kg/hr

2.48 x 106 Kglhr

1839C

0.3683 m.

0.3658 m.

FUEL ELEMENTS

*Number of Fuel Elements

*Rod Fuel Layout

*Cladding

*Fuel

*Outter clad diameter

*Clad Thickness

*Channel

400

8x8R and P8x8R

Zircaloy-2

U02

1.25 crnl 1,23 cm

0.086 cm/ 0,081 cm

Zircaloy-4

CONTROL RODS

*Number of Control Rods

*Shape

97

Cruciform

REACTOR VESSEL

Inner Diameter

Inner height

Design pressure

4.775 m.

18.447 m.

87.90 Kg/cm2

13
TABLE I I - 1 (Cont.)

RECIRCULATION SYSTEM

" Location

" Number of loops

" Loop diameter

*Nominal flow per pump

" Number of let 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.

RV'S

*Number

*Setpoint

3

74.2, 74.9, 75.6 Kg/cm2

14
TABLA I I I - 1

DESCRIPTION OF THE TRAC COMPONENTS

NUMBER COMPONENT DESCRIPTION

1 VESSEL Vessel

20 TEE Suction pipe of the recirculation loop

21 VALVE Isolation valve of the recirculation loop

22 PUMP Recirculation pump of the loop

23 VALVE Isolation valve of the recirculation loop

-24 TEE Discharge pipe of the recirculation'loop

25 JETPUMP Jet pump of the recirculation loop

31 VALVE IC Lne

32 PIPE IC Shell

40 CHAN Average bundle'

50 TEE Main steam lie from the vessel

52 TEE Main steam line from the vessel

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

71 FILL Low pressure core spray

55 TEE Main steam line to the turbine

15
TABLA IV- 1

INITIAL CONDITIONS

PARAM~ETER

Reactor Vessel

Total Core Power (100%), (Mw)

Downcomer Water Level, (in)

Steam Dome, (MPa)

Total Core Mass Flow, (kg/s)

Core Bypass , (kg/s)

Core Ap (MPa)

Recirculation Loop

Speed Pump, (rad/s)

Temperature (K)

Drive flow (m3lseg)

Feedwater System

Feedwater Mass Flow, (kgts)

MEASURED VALUE TRAC-BF1/GlJl

1380

10.53

6.96

5360

0.08

1377

10.51

6.96

5432

555

0.10

120

546

3.26

121

547

3.29

684 685

16
TABLE IV-2

CHRONOLOGY OF EVENTS

Ey=n

Closure Initiated

Scram Signal Generated

Low Water Level

Main Steam isolation Valves Closed

High pressure peak

RV's Open

RV's Closed

IZ=f (seconds)

PLANT

0.0

i.0

2

3.0-4.0

3-4

- 4

7-9

TRAC

0.0

0.3

2.8

4

5

3.5

7-9

17
TABLA V-1

RUN STATISTICS

Real lime.

CPU Time

Total number of volumes in the model

Total number of time steps

(CPU x 103)/(C x DT) = 108

RT,- 60 seconds -

CPU - 13764 seconds

C. 119

DT. 1072.

18
Dome pressure

7600000

7400000

7200000

7000000 -

Pa 6800000

6600000

6400000

6200000 *

6000000

40 50lAf 60 70 80 90

sg

Figure 11-1
Downcomer water level

11.2

11

10.8

10.6

10.4

10.2 N0

m

_____ _____ Q _____ _____ _____

__ 0 __ __ __ __

0-~

-a- -0-

10

9.8.

9.6

9.4

9.2

9

0 10 20 30 40 50 60 70 80 90

og

E Plant B 'OPlant A

Figure 11-2
Core flow

6100

6000

5900

N)'

5800

Kg/s 5700

5600

5500

5400

5300

0 10 20 30 40 s0 60 70 80 90

ag

Figure 11-3
Feedwater flow

3000 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

2500 -

2000 _____

Kg/B 1500 __ _ _U_ _ _

*1000 ______

Soo U

oK

0 1 20l 30 40 50 60 70 80 90

og

Figure 11-4
DRYWELL

SUPRESIONSUPRESSION

POOL POOL

HPCIIRV. SRV

BPV

MSIV TSV. TGV

IN,

HI-h.

/sa ics

RECIRC. PUMP

Figure III - 1
(18,8m)

(16,5m) 1 

0

*DRYER

....... TO

RELIEF VALVEg 3 4IS(D 2

1-6 1112131F1

(14,1lm)

(12,3m)

98

(I11,7m) 

7

(10,5m) 

6

I\) (8,8m)

5

(7,5m) 

4

(5,1lm) 

3

(2.4m) 

2

*1~

_T.ANDPIPE~

MIXI NGI

PLENUM!

OCaR

PLENUM

'S...

33

4.23

ISOLATION

VALVE

13 2 

FEEDWATER

2L31 9 TURBINE

JETPUM

2

1 2 3 

-1

4

1 2 

1 

2 i'ýu

RECIRCULATION

PUMP

I

(0m)

1

I

Figure 111-2 TRAC-BFI AXIAL NODALIZATION
Pressure

r%3

7600000

7400000

7200000

7000000

Pa 6800000

6600000

6400000

6200000

6000000

U

3-.--.

U

U

______ ______ U _____ ______

U- ~ ~ U U U

0 10 20 30 40 50

sec

Plant -TRAC

60 70 so 90

Figure IV-1
Mixing plenum void fraction

l's)

0.8

0.7

0.6

0.5

0.4

0.3

0.2

0.1

0

0 10 20 30 40 s0 60

sea

70.

Figure IV-2
Core flow

6100

6000

5900

.5800

IN)

Kg/s 5700

5600

5500

5400

5300

0 10" 20 30 40' 50 60 70 80 90

sec

UPlant -TRAC

Figure IV-3
Jet pump throat void fraction

0.05

0.045

0.04

0.035

0.03

0.025

0.02

0.015

0.01

0.005

0

V

0 10 20 30

sec

40 50 60

Figure IV-4
Downcomer level

12

11.5

11

m 10. 5

10

9.5

9

0 10 20 30 40 so 60 70 80

sec

90

U Plant A -TRAC KOPlant B

Figure IV-S
Pressure

7600000

7400000

7200000

7000000

Pa 6800000

6600000

6400000

6200000

6000000

00

0 10 20 30 40 50 60 70 80 90

sec

UPlant -TRAC

Level

11.2

11

10.8

10.6

10.4

m 10.2

9.8

9.6

9.4

9.2

9

0 10 20 30 40 50 60 70 80 90

sec

U Plant A -TRAC 0 Plant B

Core flow

6100

6000

5900

5800

Xgs5700

Kgu5600

5500

5400

5300

5200

0 10 20 30 40 50 60 70 80 90

sec

UPlant -TRAC

Figure IV-6

30
Pressure

7600000

7400000

7200000

7000000

Pa 6800000

6600000

6400000

6200000

6000000

MA

Emr0 10 20 30 40 50 60 -70

sec

so 90

UPlant -TRAC

Level

11.2

11

10.*8

10.6

10.4

m 10.210

9.8

9.6

9.4

9.2

9

0 10 20 30 40 s0 60 70 s0 90

sec

U Plant A -TRAC 0 Plant B

Core flow

6100

6000

5900

5800

Kg/s .5700

5600

5500

5400

5300

5200

AU

0 10 20 30 -40 50 60 70 80 90

sec

UPlant -TRAC

.Figure IV-7

31
Pressure

7600000

7400000

7200000

7000000

Pa6800000

Pa6600000

6400000

6200000

6000000

5800000

0 10 20 30 40 50 60 70

sec

EPlant -TRAC

Level

10.5

m

10

9.5

9

0 10 20 30 40 50 60 70

sec

U Plant A -TRAC KPlant B

Core flow

Kg/a

0 10 20 30 40 50 60 70

sec

UPlant -TRAC

Figure IV-8

32
Time step

0.08

0.07

0.06

0.05

sec 0.04

0.03

0.02

0.01

0

0 10 20 30

Problem see

40 50 60 7c

Figure V-i
C.
A~C FORM 335 U.S. NUCLEAR RIEGULATORY COMMISSION 1.REPORT NUMBER

1249) I A*Ouwma by NAC. Add Vol.. Sum Rev.,

I4RCM 1102. &M Add.,rdum Numbers. If ,any.) X201.3202 BIBLIOGRAPHIC DATA SHEET

(See inlstru'ctions on the ev~eruel NUREG! IA-Ol 22

2. TITLE ANO SUBTITLE 1CSP-GA-MSIV-T

Assessment of MSIV Full Closure for Santa Maria De Garofla 3. DAT REPORT PUBLISHED

Nuclear Power Plant Using TRAC-BFl (GI 31) MONTH -- EAR

June 1993

4. FIN OR GRANT NUMBER

L2245

S. AUTORIS)6. TYPE OF REPORT

J. L Crespo/University of Cantabria Technical

R. A. Ferna~ndez/Nuclenor, S. A. 7. PERIOD COVER ED Weilusjiv Dsesta

University of Cantabria

FAG Ciencias. Seccion De Fisicas. Nuclenor

c/Hernan Cortes, 26

39080 Santander, Spain

9. SPONSORING ORGANIZATION - NAMAE AND ADDRESS (It NOIC. tyPe"Sm eus abowve;fcont teem,. Provide NRC 0/vs/aon. Otfice of Reg1on. U.S. Ak#ear RegUlatorr CoMn,/ailo.

Office of Nuclear Regulatory Research

U.S. Nuclear Regulatory Commission

Washington, 13C 20555

10. SUPPLEMENTARY NOTES

11. ABSTRACT fXoo word; of 1sem

An assessment of the first 60 seconds of a spurious Main Steam Insolation Valve (MSIV's) closure for Santa

Maria Garofia Nuclear Power Plant using TRAC-BFl code is presented. Reasonable and realistic adjustments have

been made in the model to improve its performance.

This work is part of the validation set for the TRAC model that is being developed for wider use and allow to

test the code capabilities.

As a result of the analysis, it is felt that TRAC-BFl is capable of reproducing the plant behavior with an

acceptable degree of accuracy although better models are clearly needed, in addition to further noding work and

code improvements. The code took almost 14000 sec. which makes a 1/230 calculation time to real time ratio.

For this transient a mechanistic separator model is needed. It will alsq help to cut down running costs if the

vessel noding could have different number of cells at different heights. Though not very important for this transient,

the critical flow model will allow for realistic RV flow assumptions.

There are not guidelines available for separator modelling in transients. It has been found that a detailed noding

in the separator region may be needed to represent steamn-water interaction.

12. KEY WORDSIDESCRIPTORS (Lit woidsor ophrm tai~e, v/Uailr~atimheas In be sing the rnort.) 13. AVAILABILITY STATEMENT

ICAP Unlimited

Santa Maria De Garofta 14. SECURITY CLASSIFICATION

MSIV I'Thit Page)

TRAC Unclassified

(Tilis Pecan,

Unclassified

15. NUMBER OF PAGES

16. PRICE

?aRC FORM 335 (2491
Federal Recycling Program

NUREGAA-0122 NURE/IA-122ASSESSMENT OF MSIV FULL CLOSURE FOR SANTA MARIA

DE GARONA NUCLEAR POWER PLANT USING TRAC.BF1 (Glui)

JUNE 1993

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

FIRST CLASS MAIL

POSTAGE AND FEES PAID

USNRC

PERMIT NO. G-67

OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300