... 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