Radiation Shielding Against Neutron Using Different Fillers: Review Study

Investigation of the IAEA benchmark problem using MCNP code

SHIELDING ASSESSMENT OF THE ET:RR-1 REACTOR UNDER POWER UPGRADING.

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

SHIELDING ELECTRONIC CIRCUITS FROM SPACE-NEUTRONS AND CAPTURE GAMMA-RAYS

- Beam Neutron Optimization for Boron Neutron Capture Therapy (BNCT) facility

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

Evaluation of gamma rays shielding competence for bentonite clay /PVA polymer matrix using MCNPX code

A CONCEPTUAL GAMMA SHIELD DESIGN USING THE DRP MODEL COMPUTATION

Proposed New Different Coincidence Neutron Detection Systems using Monte Carlo Simulation.