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... Reactors (ASRR-II) held at Serpong, Indonesia. 23-25 May 1989. S. Ganesan. IAEA NUCLEAR DATA SECTION, WAGRAMERSTRASSE 5, A-1400 VIENNA ...
INDC(INS)-001 Distri. L

International Atomic Energy Agency

I N

D O

INTERNATIONAL

NUCLEAR

DATA

COMMITTEE

/

PLANS FOR USE OF ENDF/B IN REACTOR RESEARCH IN INDONESIA

Budi Santoso, Ahmad Syaukat, Iyos Subki Badan Tenaga Atom Nasional Jl. K.H. Abdul Rrfchim, Kuningan Barat Mampang Prapatan, P.O. Box 85 kby Jakarta 12710, Indonesia

S. Ganesan Indira Gandhi Centre for Atomic Research Kalpakkam 603 102, India

Paper presented in the second Asian Symposium on Research Reactors (ASRR-II) held at Serpong, Indonesia 23-25 May 1989

IAEA NUCLEAR DATA SECTION, W A G R A M E R S T R A S S E 5, A-1400 VIENNA

INDC(INS)-001 Distri. L

PLANS FOR USE OF ENDF/B IN REACTOR RESEARCH IN INDONESIA

Budi Santoso, Ahmad Syaukat, Iyos Subki Badan Tenaga Atom Nasional Jl. K.H. Abdul Rochim, Kuningan Barai Mampang Prapatan, P.O. Box 85 kby Jakarta 12710, Indonesia

S. Ganesan Indira Gandhi Centre for Atomic Research Kalpakkam 603 102, India

Paper presented in the second Asian Symposium on Research Reactors (ASRR-II) held at Serpong, Indonesia 23-25 May 1989

Reproduced by the IAEA in Austria July 1989 89-03302

PLANS FOR USE OF ENDF/B IN REACTOR RESEARCH IN INDONESIA Budi Santoso, Ahmad Syaukat, Iyos Subki BADAN TENAGA ATOM NASIONAL. Jl KH. Abdul Rochini, Kuningan Barat. Mampang Prapatan. P. 0. Box 85 kby, Jakarta 12710, Indonesia S. Ganesan Indira Gandhi Centre for Atomic Research, Kalpakkam 603102 India

Abstract f Nuclear

data

are

numerical

constants

of

nature

which

quantify the nuclear behaviour of all elements and isotopes which make

up

needed

the reactor medium and its environment,

and which

are

as input for performing design calculations for safe

and

reliable

operation

specially

computerized files such as the Evaluated Nuclear known as ENDF/B.

recommended

values

formatted

briefly

of

in

are

Data

The development of base technology in

scheme of original reactor design

mastering

and

of

data

in

the

form

The nuclear

available

File-B,

the

of nuclear reactors.

the art of ENDF/B

calculations involves the

data

processing.

This

paper

discusses the current status of this activity in Jakarta

gives an account of the future plans,

role of ENDF/B in reactor

with emphasis on

the

calculations.

I. Introduction In

order to perform

starting any

rigorously all

neutronic

calculations

from the basic evaluated data file such as ENDF/B,

research or original designs,

initiated

in BATAN,

Indonesia.

a long,term project has The project involves

for been

starting

from basic data file, mastering the art of ENDF/B data processing for

reactor

ENDF/B

in

applications original reactor

and creating the design

capability

calculations;

to

This,

use paper

briefly

discusses

the

current

status

of

this

activity

and

gives an. account of the future plans. Basic

evaluated nuclear data libraries such as ENDF/B,

JENDL-2, BROAND are.available nuclear

data

evaluated

in ENDF/B format from

centres such as IAEA,

Vienna.

The

international use

nuclear data libraries in research reactor

calculations creating

base

technology

for indigenous

and power reactors.

and

for

the

design of reactors including

NEA Data Bank,

possible to readily

research

distribution

adapt these multigroup

data

unless

imported.

Most

is that the use of such condensed

libraries

continue dependence on foreign efforts in updating the

libraries

and

drawback

local

RSIC(USA) etc., it has not

compatible neutronic code systems are also

important

for

Although condensed multi-group libraries and

centres such as IAEA,

the

application

enhancing

few group constants are available from international

been

such

has been realized to be natural starting point

the

capability

of

the

possibility

calculations

in

a

of

detailed

rigorously manner

performing starting

from

evaluated data file is ruled out in such cases. to

use

the

ENDF/B formatted basic

libraries

section

for

formulated and presented Indonesian

Nuclear

ENDF/B

outlined.

are

progress BATAN

neutron

reactor design in in this paper.

Energy The

Programme present

sensitivity

presents

Indonesia

4

GROUPIE

etc.,

been

The advantages for in

the

paper

direct

also

use

outlines system to

the of the in

further design

The experience gained thus far, in BATAN

the use of IAEA pre-processing codes LINEAR,

SIGMAl,

cross

have

consolidate the routine use of ENDF/B in research reactor

in

plans

interaction

details of proposed activities

calculations in BATAN.

basic

Therefore

made thus far in the installation of ENDF/B

and

the

has

been very useful

RECENT, in

FIXUP,

understanding

several aspects of ENDF/B processing for reactor applications

and

in appreciating the further efforts needed

use

of

ENDF/B

system

Appreciation constants

of

in

situations

physical

existing

principles and

using the IAEA pre-processing

FEDGROUP-C, developed

REXl,

REX2

for the successful

and

REX3

Indonesia;

generation

of

or use of

NJOY

code

processing

ENDF/B formatted

libraries such as ENDF/B, available

from IAEA.

As

understanding creation

JENDL-2,

basic

cross

evaluated

of

and

of the

strategy basic

optical

term

process

As an example to cite

here,

already been adapted on IBM-PC/AT in

In reactor design,

Data in Reactor

adequate

used

the

in

in BATAN

neutron and time. integro-dif f e'rential gives

user-friendly statistical user-friendly

BATAN.

Design

given a set of boundary conditions,

the

one

density

direction of motion and energy of

the

There are seven independent variables in

the

equation

the transport equation

describes

to

interaction

at solving the neutron balance equation for neutron function of position,

IBM-

and

the

is a well known

are

planning

work has been initiated

II.Role of Nuclear

as

long

evaluation

model program ABAREX which

has

aims

for

perform nuclear model based neutron

codes.

data

BROAND, ENDL etc., which

section calculations using well established

computer

code

part

it is

These projects can make use of both

of ENDF/B itself,

understand

the

important

PC/AT personal computers and the VAX 8550 computer with memory.

sytem

As part of nuclear data processing activities

by

as

team to realize

planned to take up intercomparison of nuclear data for isotopes

group

codes and codes such

in USA need a long term committed

objectives.

in

neutron

that one has to solve. (Bell and

population

Glasstone,

density

in the

Table

1970) form

1

which of

an

5

Table 1

The seven dimensional balance equation for the neutron in a nuclear reactor

The basic neutron balance equation for nuclear reactors in a volume element dV for neutron having energy between E and E+dE and flight direction in solid angleJLA. around jjis given by:

Bn

E , .2 , t)

(r,

- -

JT.VN

r>t

(r,

E,a,

t)v

Leakage

E sl'

(?,

E,jt,

t)

Loss due to absorption and scattering

e ) j x ' , t)v' dE'

) N (r,

£

-

dSil

source due to scattering from other directions and energies

+ Y

(E ) \ ( 4

m'ILI) N U ,

e'a,

t)

V ( E ' ) dE' d A '

/\

Fission source

+ S (r, E,-n-, t) External

Symbols v N

source

: Neutron speed corresponding to energy E : Neutron angular density : Total neutron cross section

nt S "r E t 2 . (E1 —> E, J1-' s ' (E) r^j

: : : : :

External neutron source space coordinate Unit vector in the direction of neutron Energy of the neutron time

) : Scattering cross section for transfer of neutron from (E 1 , Jt' ) into : Fission neutron spectrum : Average number of neutrons produced per Fission Note that the macroscopic cross sections depend on space ^ , time t and energy E

Boundary conditions : —»

: : :

6

N (r, E, -n. , t) = 0 For energies greater than maximum of source energy (15 MeV) Continuity at the interface between two media N (r, E, .A. , t) = 0 At the outer free surface for directions entering the system

integro-differential equation the

Physicists

try to

solve

which is the starting point for the physics

reactor,

materials to

equation.

by

taking

all

the

nuclear

this

design

properties

of

and the geometry of the reactor into account in

obtain

the important characteristics of the nuclear

such as the K-eff,

namely,

of the

order reactor

the effective multiplication

factor,

critical mass, fuel enrichment, neutron flux, Doppler and coolant void

reactivity

coefficients,

dynamic

characteristics

normal and abnormal operating conditions etc. of the

during

The physics design

the reactor forms part of the overall engineering

design

reactor which includes the thermal hydraulics design of

of the

The macroscopic cross sections needed as input to solve

system.

the transport equation in Table 1 are obtained from the specified atom

densities

of

individual

isotopes

and

the

numerical values of microscopic neutron-nuclear section

recommended

interaction cross

data of individual isotopes which are available

evaluated

nuclear

computers

force

approximation multigroup

data files. reactor

The limitation of the

physicists to

to solve the balance equation.

form of the balance equation,

sections

are

sections

of

invoke

to

be

generated.

individual data

isotopes

The

energy

files are condensed

the

existing multigroup

For solving

the

which are

the

in

the

multigroup

cross

dependent

cross

available

evaluated

nuclear

into

a

constants

through averaging procedures by following the

in

few

the group

law

conservation of reaction rates for each of the energy bins. process

of

condensation is generally performed in a few

The basic file is pre-processed neutronic

calculations

in

of The

steps.

into 2000 fine groups followed by

2000

groups

to

get

the

spectrum

Characterising the homogeneous composition of each reagion of the reactor.

The- cross;'sections are then collapsed to 200 groups

at

7

Fig. 1: Various stages/steps in the nuclear reactor

which

level

the

heterogeneity of the lattice may

calculations

be

treated.

With further neutronic calculations, condensation is performed to generate

a

few

consideration.

froup

Fig.

1

constants shows,

stages in the nuclear reactor

One WIMS

code

for

the

reactor

in a general way,

various

calculations.

of the well known codes developed i n U. which

the

under

can be used f o r a

wide

spectrum

K.

is

the

of

reactor

systems.

Gaining

sufficient experience with such code

systems

gives the necessary expertise that will enable Serpong to prepare and

generate

data

from

reactor calculations.

ENDF/B for light

water

and

research

The use of FEDGROUP-C code which

processes

ENDF/B and produces results of multigroup cross sections in format A.

or the use of the more general NJOY code system of U.

which

also has the option to produce results in WIMS

could

be taken up in Serpong.

such

code

systems

computational

reactor

demand

a

physics

thorough

knowledge

when one takes up

III.

With plotter,

the

of

the

S.

format

The understanding of the use

practical thermal power and research reactor

PCs,

WIMS

of

whole task

of

calculations.

Implementation

available facility like

VAX-8550,

Packard

Microvax Workstation II/GPX, laser printer and

X-Y

desk-top

the implementation of data base development project will be

carried out in steps. a seven weeks mission January

At the request of BATAN,

IAEA

(Dr. S. Ganesan) from 19 November

undertook 1988 to 6

1989 following a one week preparatory mission which

was

carried out from 24th September to 2nd October 1988 by Mr. Andrej Trkov. The

following capabilty as a result of this mission

presently The

exists

in Jakarta.

team .in Jakarta can retrieve and understand

the

contents

of ENDF/B formattted data. The team knows how to use the manuals. The

team

in

Jakarta

has acquired the

confidence

capable of correctly using the following pre-processing

and

are

codes.

LINEAR: To linearize ENDF/B data RECENT: To reconstruct resonance cross

sections

9

SIGMAl:

To

obtain

Doppler broadened cross sections

at

higher

temperatures. MERGER:

To

obtain

ENDF/B

tape

for

specific

isotopes

and

reactions. MIXER:

To obtain total cross sections of mixture

VIRGIN: To study and interpret self-transmission FIXUP:

To correct for negative cross

sections.

LEGEND: To correct for negative angular GROUPIE:To

obtain

factors

which

total,

elastic,

experiments

distributions.

multigroup cross sections and

are

temperature and fission,

self-shielding

composition

scattering

and

Bondarenko definition; to obtain multi-band

dependent

capture,

based

for on

parameters.

It is planned that the programs EVALPLOT and COMPLOT will be commissioned connected master

as

the hardware i.

to mainframe computer.

the X-Y

plotter

As an exercise to

will

and

Mr.

a team consisting of

Karsono

and led by

S.

Mr.

immediate

Bunjamin,

Ganesan

The listing of this program

Mr.

successfully

developed a code for plotting any desired cross section from 3 of ENDF/B.

be

thoroughly

the retrieval of ENDF/B data and to satisfy the

plotting requirements, Dewanto

e.

is available from

file the