Modeling and neutronic analysis of pin-cell comprising nuclear fuel with different chemical composition and neutron moderator using Monte Carlo code OpenMC

Highlights: • Pin-cell analysis of eight $\left(08\right)$ models using the Monte-Carlo code OpenMC. • $k_{\mathrm{eff}}$ was calculated and validated by the deterministic code Dragon. • Depletion and production analysis of different isotopes were carried out. • Variations of $k_{\mathrm{eff}}$ with enrichment, Moderator/Fuel ratio were studied. • Effect of boron concentration on $k_{\mathrm{eff}}$ was also presented. Selection of a perfect nuclear reactor fuel demands the neutronic analysis of the available fuel materials in particular reactor environment. This paper summarizes the pin-cell analysis of eight $\left(08\right)$ models considering five $\left(05\right)$ different chemical compositions of nuclear fuel, three (03) different neutron moderator and two (02) types of cladding materials using the open-source Monte-Carlo code OpenMC. The fuel element includes UO₂, UN, UC₂, MOX, and Thorium fuel, whereas light water, heavy water, and graphite were considered as moderator. The effective multiplication factor ($k_{\mathrm{eff}}$) was calculated for all the models and validated by the deterministic code Dragon. It was found that, while keeping the dimensions fixed, the graphite and heavy water moderated fuel pin-cells did not reach criticality. So, an appropriate pitch was found where these two cells reached criticality. Depletion and production of different fissionable materials, as well as neutron poison, were studied. Effect of enrichment, moderator to fuel ratio, and boron concentration on the $k_{\mathrm{eff}}$ in a typical water-moderated UO₂ with zircalloy-4 clad pin-cell were also presented.