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

• Published in: Pramāṇa 2013, Vol.80 (1), p.105-120
• Main Authors: FEGHHI, SEYED AMIR HOSSEIN, REZAZADEH, MARZIEH, KADI, YACINE, TENREIRO, CLAUDIO, AREF, MORTEZA, GHOLAMZADEH, ZOHREH
• Format: Article
• Language: English
• Subjects: Astronomy; Astrophysics and Astroparticles; Observations and Techniques; Original Article; Physics; Physics and Astronomy
• ISSN: 0304-4289; 0973-7111
• DOI: 10.1007/s12043-012-0462-7

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 233 U converted from 232 Th 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.