ADVANCED STUDIES ON PLUTONIUM FOR PU ISOTOPE TRANSMUTATION AND PU SEPARATION

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

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

Using Thorium Based Fuel in AP1000: Steady State Analysis

Characterization of Pu-bearing materials by non-destructive spectrometric approach: relevance for nuclear forensics

VERIFICATION OF 235U, 238U BY MATHEMATICAL CALIBRATION CURVE FOR POINT SOURCES BY USING MONTE CARLO

Isotopic Inventory and Activity Calculations of Fukushima Daiichi Unit-1 Accident

Determination of Uranium Isotopic Ratios by Alpha Spectrometry for Nuclear Materials Control Purposes

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