Graphite moderated 252Cf source

• A Monte Carlo study of a 252Cf source moderated by graphite was carried out.
• The neutron spectra and the total fluence in the irradiation channel was estimated.
• The total neutron fluence distribution was estimated in the array.
• The H⁎(10) distribution was estimated in the array.

The Thorium molten-salt reactor is an attractive and affordable nuclear power option for developing countries with insufficient infrastructure and limited technological capability. In the aim of personnel training and experience gathering at the Universidad Simon Bolivar there is in progress a project of developing a subcritical thorium liquid-fuel reactor. The neutron source to run this subcritical reactor is a 252Cf source and the reactor will use high-purity graphite as moderator. Using the MCNP5 code the neutron spectra of the 252Cf in the center of the graphite moderator has been estimated along the channel where the liquid thorium salt will be inserted; also the ambient dose equivalent due to the source has been determined around the moderator.