Simulation of BEAVRS Benchmark at Hot Zero Power Using MCNP6

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

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

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

Design of portable shield for neutron sources using MCNP Computer Code

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

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

Development of Two Software Packages for Connecting with the MCNPX and NJOY Nuclear Codes to Directly Obtain and Extract Multi-Group Nuclear Cross Sections for Any Isotope

Monte carlo method estimates for internal dosimetry of well differentiated thyroid cancer patients

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