Increasing demands for better perfoming glasses have lead to current investigating of the sturctural properties of glasses for optimum performances. Calcium sulphate borophosphate glasses of different compositions were prepared using melt quenching technique. The glass forming ability and stability were checked using Differential thermal analyzer (DTA). Density and molar volume had been evaluated and analyzed. From the results of XRD, the absent of discrete and continuous sharp peaks confirmed the amorphous nature of the glass compositions while the results from both IR and Raman revealed the existence of SO4, BO4, BO3, P-O-P and PO43-. Addition of CaSO4 to borophosphate influenced the conversion of the dominant BO3 groups to BO4 groups. The structure of the samples was mainly based on metaphosphate, diphosphate and BO4 units, which became depolymerized with addition of CaSO4 content. The glass forming ability and thermal stability were found to increase with an increase in the concentration of modifier content. Glass density and molar volume is found to be between 2.146 to 2.314 gcm-3 and 45.794 to 48.880 m3mol-1 respectively. It is observed that the density of glass increased while the molar volume also increased with respect to increase in concentration of CaSO4 in the glass compositions. We analysed our data using different mechanisms and compared the results with previous works. Our findings show that this glass could be beneficial and considered as a good candidate for optical devices applications.

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The International Atomic Energy Agency (IAEA) requires that all test and research reactors operating on Higher Enriched Uranium (HEU) should be converted to Low Enriched Uranium (LEU) for safety and security purposes. Nigeria having a Miniature Neutron Source Reactor (MNSR) has been long interested in fuel technological research not just to develop the area but also to meet with resolution on the nuclear treaty set out by the global nuclear regulatory body. In this study, reactor kinetic parameters such as effective delayed neutron fractions, prompt neutron lifetime as well as mean neutron generation time were analysed for Nigerian Research Reactor-1 (NIRR-1). Serpent Monte Carlo code 1.17 is used in the analysis. For delayed neutron parameters determination, we used fission probability iteration under one averaged generation time and neutron population rate.The calculated values for delayed neutron were recorded as analogue prompt and implicit prompt neutron lifetime, reproduction time and emission time are in the order of 3×10-7 (s), in agreement with the calculated data from the nuclear data libraries and some literature.The result for delay neutron fraction and other time-based parameters support the fuel core conversion for NIRR-1.The computational and pictorial results obtained from Serpent code simulation described well the transient behavior of the delayed neutron in this reactor.The analytical results also spelled out the relevance and compatibility of low enriched uranium dioxide fuel over higher enriched type.The result of this study conforms with other results obtained from similar reactors but with different Monte Carlo codes and with higher enriched uranium