Highlights: ► Irradiation site with graphite surrounding it was design and modelled with MCNP-5. ► Thermal neutron flux in the irradiation channel increased with presence of graphite. ► The thermal neutron flux gradient decreased along the diameter of the tube. ► Improved thermal neutron distribution in the channel making the site apt for LSNAA. ► The simulation results agree very well with available data such as final keff = 1.0039. – Abstract: Monte Carlo N-Particle code (MCNP-5) was employed to simulate the neutron flux profile in a newly designed irradiation site surrounded by a graphite thermal column for Large Sample Neutron Activation Analysis, in the Ghana Research Reactor-1. The results shows that the average thermal neutron flux in the irradiation channel surrounded with 6 cm thick graphite was 5.45 × 1010 n cm−2 s−1 as compared with 1.74 × 1010 n cm−2 s−1 when there is no graphite around the irradiation channel. This shows an increase in the thermal neutron flux in the irradiation channel by a factor of 3.13. The thermal neutron flux gradient decreased at 7.0 × 109 n cm−2 s−1 per cm to 3.90 × 107 n cm−2 s−1 per cm when the channel is surrounded by 6 cm thick graphite which makes it very suitable for Large Sample Neutron Activation Analysis. The simulation agrees very well with the experih the experimental keff of 1.00400 as compared with the final keff of 1.00390 recorded by this simulation.
Source/Report: Annals of Nuclear Energy (Oxford); v. 42; ISSN 0306-4549; ; CODEN ANENDJ; Apr 2012; p. 131-134; S0306-4549(11)00447-6; Available from http://dx.doi.org/10.1016/j.anucene.2011.11.019; Copyright (c) 2011 Elsevier Science B.V., Amsterdam, The Netherlands, All rights reserved.; Country of input: International Atomic Energy Agency (IAEA)
Publ. Year: 2012