• Morteza Aref

      Articles written in Pramana – Journal of Physics

    • Neutronic simulation of a research reactor core of (232Th, ${}^{235}$U)O2 fuel using MCNPX2.6 code

      Seyed Amir Hossein Feghhi Marzieh Rezazadeh Yachine Kadi Claudio Tenreiro Morteza Aref Zohreh Gholamzadeh

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      The small reactor design for the remote and less developed areas of the user countries should have simple features in view of the lack of infra-structure and resources. Many researchers consider long core life with no on-site refuelling activity as a primary feature for the small reactor design. Long core life can be achieved by enhancing internal conversion rate of fertile to fissile materials. For that purpose, thorium cycle can be adopted because a high fissile production rate of 233U converted from 232Th can be expected in the thermal energy region. A simple nuclear reactor core arranged 19 assemblies in hexagonal structure, using thorium-based fuel and heavy water as coolant and moderator was simulated using MCNPX2.6 code, aiming an optimized critical assembly. Optimized reflector thickness and gap between assemblies were determined to achieve minimum neutron leakage and void reactivity. The result was a more compact core, where assemblies were designed having 19-fuel pins in 1.25 pitch-to-diameter ratio. Optimum reflector thickness of 15 cm resulted in minimal neutron leakage in view of economic limitations. A 0.5 cm gap between assembles achieved more safety and $2.2$% enrichment requirements. The present feasibility study suggests a thermal core of acceptable neutronic parameters to achieve a simple and safe core.

    • A benchmark study on uncertainty of ALICE ASH 1.0, TALYS 1.0 and MCNPX 2.6 codes to estimate production yield of accelerator-based radioisotopes

      Seyed Amirhossen Feghi Zohreh Gholamzadeh Zahra Alipoor Akram Zali Mahdi Joharifard Morteza Aref Claudio Tenreiro

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      Radioisotopes find very important applications in various sectors of economic significance and their production is an important activity of many national programmes. Some deterministic codes such as ALICE ASH 1.0 and TALYS 1.0 are extensively used to calculate the yield of a radioisotope via numerical integral over the calculated cross-sections. MCNPX 2.6 stochastic code is more interesting among the other Monte Carlo-based computational codes for accessibility of different intranuclear cascade physical models to calculate the yield using experiment-based cross-sections. A benchmark study has been proposed to determine the codes' uncertainty in such calculations. ${}^{109}$Cd, ${}^{86}$Y and ${}^{85}$Sr production yields by proton irradiation of silver, rubidium chloride and strontium carbonate targets are studied. $^{109}$Cd, $^{86}$Y and $^{85}$Sr cross-sections are calculated using ALICE ASH 1.0 and TALYS 1.0 codes. The evaluated yields are compared with the experimental yields. The targets are modelled using MCNPX 2.6 code. The production yields are calculated using the available physical models of the code. The study shows acceptable relative discrepancies between theoretical and experimental results. Minimum relative discrepancy between experimental and theoretical yields is achievable using ISABEL intranuclear model in most of the targets simulated by MCNPX 2.6. The stochastic code utilization can be suggested for calculating $^{109}$Cd, $^{86}$Y and $^{85}$Sr production yields. It results in more valid data than TALYS 1.0 and ALICE ASH 1.0 in noticeably less average relative discrepancies.

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