RRFM 2009 Transactions - European Nuclear Society

analysis procedure (ENAA) except via the use of movable thermal neutron shields. In this
work, in order to eliminate exposure to radiation dose from movable shields and perform
activation within a permanently fixed Cd shield, a conceptual permanent Cd lined inside
a large outer irradiation of NIRR-1 was simulated using Monte Carlo transport code,
MCNP5 version 1.40. Furthermore, results of calculated data of Cd-ratios together with
advantage factors for the determination of I, Br, Cl and K in one of the outer irradiation
channels of NIRR-1 in geological samples are presented.
Materials and methods
The Monte Carlo N-Particle (MCNP) code and a set of neutron cross-section data were
used to develop an accurate three-dimensional computational model of current NIRR-1
core fueled with HEU (Jonah et al., 2007). The geometry of the reactor core was modeled
as closely as possible including the details of all the fuel elements, reactivity regulators,
the control rod, all irradiation channels, and Be reflectors. The MCNP code is a generalpurpose
code for the transport of neutrons, photons and electrons (Breimester, 2000). The
neutron energy regime is from 10 -11 MeV to 20 MeV for all isotopes and up to 150 MeV
for some isotopes, the photon energy regime is from 1 keV to 100 GeV, and the electron
energy regime is from 1 keV to 1 GeV. The simulation of radiation transport in matter
involves the tracking of particles according to established probabilistic laws, commonly
known as cross sections. An important feature of the MNCP is capability to calculate k ef f
eigen values for fissile systems. Like input wase executed as KCODE source problem for
criticality calculations using the MCNP5 code version 1.40, which runs on a Linux
Cluster in the Computational Reactor Physics laboratory at the Centre for Energy
Research and Training, Ahmadu Bello University, Zaria. The Linux Cluster consists of
four personal computers, each with 512 MHz processor speed and 1024 M-Byte RAM. It
was executed on multiple processors using MPI parallel computing capabilities with
MCNP5 executables. All jobs were executed with the “TASK n” option made with
100,000 particles in 400 cycles. One of the basic tally card f4:N, the track length estimate
in the cell with different energy bins was constructed for the calculation of axial neutron
flux distributions in the irradiation channels according to the 640 group energy structure.
An MCNP diagram of NIRR-1 showing a Cd-liner in one of the large outer irradiation
channels is depicted in Fig. 1.
According to the Hogdahl convention the Cd-ratio of nuclides of interest can be
calculated from the expression given in equation 1.
R
CD
f
= + 1
Q ( α)
o
(1)
Where,
f is the ratio of thermal-to-epithermal flux ratio determined experimentally,
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