Analysis of Small Modular SMART Reactor Core Fuel Burn up using MCNPX code

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

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

Investigation of the IAEA benchmark problem using MCNP code

Investigation of Thermal Hydraulic Behavior of Dry Spent Fuel Cask Under Forced Convection

Using Thorium Based Fuel in AP1000: Steady State Analysis

Simulation of BEAVRS Benchmark at Hot Zero Power Using MCNP6

Analysis of Optimum Accident Tolerant Fuel and Cladding Behavior in Advanced Pressurized Water Reactor

Research Reactor Safety Analysis under Partial Loss of Flow without SCRAM