Neutronics and Dose Calculation for Prospective Spent Nuclear Fuel Cask for Ghana Research Reactor-1 Facility.

Abstract

Ghana Research Reactor-1 core is to be converted from Highly Enrich Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel in the near future; a storage cask will be needed to store the HEU fuel. Notwithstanding the core conversion process, it is also important for the facility to have a storage cask ready when the fuel is finally spent to temporarily store the fuel until permanent storage is provided. Winfrith Improved Multigroup Scheme-Argonne National Laboratory (WIMS-ANL), REactor BUrnup System (REBUS), Oak Ridge Isotope GENeration (ORIGEN2) and Monte Carlo ―N‖ Particle (MCNP5) codes have been used to design the cask. WIMS-ANL was used in generating cross sections for the REBUS code which was used in the burnup calculations. The REBUS code was used to estimate the core life time. An estimated core life of approximately 750 full-power-equivalent-days was obtained for reactor operation of 2hours a day, 4 days a week and 48 weeks in a year. The ORIGEN2 code recorded U-235 burnup weight percent of 2.90% whilst the result from the REBUS3 code was 2.86%. The amount of Pu-239 at the end of the irradiation period was 145 mg which is very low relative to other low power reactors. Isotopic inventory obtained from the ORIGEN2 and REBUS3 runs were used in setting up the MCNP5 input deck for the MCNP5 calculation of the criticality and dose rate. Six cask design options were investigated. The materials for the casks designs were selected bas on their attenuation coefficient properties and their high removal cross section properties. The various materials were arranged in no specific order in multilayered casks. The reason for investigating six casks was to look at various arrangements of the cask layers that will optimize effective shielding. The spent nuclear fuel at discharge was used as the

radioactivity source during the MCNP simulation. The multilayer cask shield comprise of serpentine concrete of density 5.14 g/cm3 and thickness 21.94cm which was used as the main gamma shield, lead (two thick layers of 1.91cm and 2.50cm respectively), boron carbide (two layers of 2cm thick each), resin (2cm thick), stainless steel (two thick layers of 2.63cm and 1.17cm respectively) and aluminium (2cm thick). Serpentine was chosen because it has higher water content than ordinary concrete thereby helping in neutron shielding. The casks were designed to be cooled by natural circulation and to have radii of approximately 60cm hence making them relatively portable. Effective multiplication factor values of Cask A, Cask B, Cask C, Cask D, Cask E and Cask F were recorded as 0.02969±0.00001, 0.06304 ± 0.00002, 0.19809 ± 0.00027, 0.15393 ± 0.00025, 0.02028 ± 0.00001 and 0.01717 ± 0.00001 respectively. This showed that all the six designs were capable of keeping the spent fuel sub critical.