Criticality Safety Analysis for Cask Design with High Discharged Fuel Burnup Using MCNPX Code

Analysis of Fuel Burn-up Calculations of 3600 MWth Sodium cooled Fast Reactor Core

The MOX Fuel Performance Analysis of ALFRED Core Using MCNPX Transport Code

Estimation of the Reactivity Temperature Coefficients of the AP-1000 Reactor by MCNPX Code

Using Thorium Based Fuel in AP1000: Steady State Analysis

Core calculations of the Molten Salt Fast Reactor using different computational codes

Simulation of BEAVRS Benchmark at Hot Zero Power Using MCNP6

Investigation of the IAEA benchmark problem using MCNP code

Core Physics Analysis of An Assembly of HTTR Reactor using Homogeneous and heterogeneous model.

Analysis of Thorium Performance in Lead cooled Fast Reactor