Evaluation of ENDF/B-VI.8 and ENDF/B-VII.0 Nuclear Data Libraries using OSMOSE Samples of MINERVE Reactor

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

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

Investigation of the IAEA benchmark problem using MCNP code

Simulation of BEAVRS Benchmark at Hot Zero Power Using MCNP6

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

Determination of Temperature Feedback Coefficients for OPR-1000 Fuel Assembly Using Adjusted Cross Section Libraries

Control Rod Worth Uncertainty Propagation from Nuclear Data for NUR Research Reactor

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