R.B.M. Sogbadji

Characterization of low power research reactor neutrons for the validation of k0-INAA standardization based on k0-IAEA software

Channel Bsite2 of Ghana research reactor-1 has been characterized for k0-INAA application. Cadmium ratio and bare multi-monitor were used to determine flux parameters using 0.1%Au-Al, Fe, and Zr wire as flux monitors. The parameters determined were 18.36±1.91, 0.0479±0.012, 5.12×10(11)±0.42×10(11) ncm(-2)s(-1), 2.74×10(10)±0.14×10(10) ncm(-2)s(-1), 7.73×10(10)±0.16×10(10) ncm(-2)s(-1) and 16.75±1.58, -0.034±0.0028, 4.28×10(11)±1.71×10(11) ncm(-2)s(-1), 2.55×10(10)±0.15×10(10) ncm(-2)s(-1) respectively for thermal-to-epithermal flux ratio, alpha, thermal neutron, epithermal neutron and fast neutron flux using cadmium ratio and multi-monitor method accordingly. The k0-INAA performance assessment based on z-score distributions showed most results within |z|<2.

Design of boron carbide-shielded irradiation channel of the outer irradiation channel of the Ghana Research Reactor-1 using MCNP

The MCNP model for the Ghana Research Reactor-1 was redesigned to incorporate a boron carbide-shielded irradiation channel in one of the outer irradiation channels. Extensive investigations were made before arriving at the final design of only one boron carbide covered outer irradiation channel; as all the other designs that were considered did not give desirable results of neutronic performance. The concept of redesigning a new MCNP model, which has a boron carbide-shielded channel is to equip the Ghana Research Reactor-1 with the means of performing efficient epithermal neutron activation analysis. After the simulation, a comparison of the results from the original MCNP model for the Ghana Research Reactor-1 and the new redesigned model of the boron carbide shielded channel was made. The final effective criticality of the original MCNP model for the GHARR-1 was recorded as 1.00402 while that of the new boron carbide designed model was recorded as 1.00282. Also, a final prompt neutron lifetime of 1.5245 × 10(-4)s was recorded for the new boron carbide designed model while a value of 1.5571 × 10(-7)s was recorded for the original MCNP design of the GHARR-1.

Neutronic Study of Burnup, Radiotoxicity, Decay Heat and Basic Safety Parameters of Mono-Recycling of Americium in French Pressurised Water Reactors

The reprocessing of actinides with long half-life has been non-existent except for plutonium (Pu). This work looks at reducing the actinides inventory nuclear fuel waste meant for permanent disposal. The uranium oxide fuel (UOX) assembly, as in the open cycle system, was designed to reach a burnup of 46GWd/T and 68GWd/T using the MURE code. The MURE code is based on the coupling of a static Monte Carlo code and the calculation of the evolution of the fuel during irradiation and cooling periods. The MURE code has been used to address two different questions concerning the mono-recycling of americium (Am) in present French pressurised water reactors (PWR). These are reduction of americium in the clear fuel cycle and the safe quantity of americium that can be introduced into mixed oxide (MOX) as fuel. The spent UOX was reprocessed to fabricate MOX assemblies, by the extraction of plutonium and addition of depleted uranium to reach burnups of 46GWd/T and 68GWd/T, taking into account various cooling times of the spent UOX assembly in the repository. The effect of cooling time on burnup and radiotoxicity was then ascertained. After 30 years of cooling in the repository, the spent UOX fuel required a higher concentration of Pu to be reprocessed into MOX fuel due to the decay of Pu-241. Americium, with a mean half-life of 432 years, has a high radiotoxicity level, high mid-term residual heat and is a precursor for other long-lived isotopes. An innovative strategy would be to reprocess not only the plutonium from the UOX spent fuel but also the americium isotopes, which presently dominate the radiotoxicity of waste. The mono-recycling of Am is not a definitive solution because the once-through MOX cycle transmutation of Am in a PWR is not enough to destroy all americium. The main objective is to propose a ‘waiting strategy’ for both Am and Pu in the spent fuel so that they can be made available for further transmutation strategies. The MOX and americium isotopes (MOXAm) fuel was fabricated to see the effect of americium in MOX fuel on the burnup, neutronic behaviour and radiotoxicity. The MOXAm fuel showed relatively good indicators on both burnup and radiotoxicity. A 68GWd/T MOX assembly produced from a reprocessed fuel spent 46GWd/T UOX assembly showed a decrease in radiotoxicity as compared with the open cycle. All fuel types understudied in the PWR cycle showed a good safety inherent feature with the exception of some MOXAm assemblies that have a positive void coefficient in specific configurations, which would not be consistent with safety features. DOI: http://dx.doi.org/10.5755/j01.erem.72.3.12790