Recent advances in Monte Carlo burnup calculations for PWR used fuel libraries
Abstract Not Provided
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Engineering topics
Fensin, Michael L
Publications and source records attributed to Fensin, Michael L.
A method for computing fixed-source driven subcritical multiplication using batch mode in MCNPX
Abstract Not Provided
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
The role of Monte Carlo burnup calculations in quantifying plutonium mass in spent fuel assemblies with non-destructive assay
Abstract Not Provided
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
The role of Monte Carlo burnup calculations in quantifying plutonium mass in spent fuel assemblies with non-destructive assay
Abstract Not Provided
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
MCNPX Simulation of a Passive Gamma System to be Used in a Spent Fuel Plutonium Assay Strategy
Abstract Not Provided
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Description of the spent nuclear fuel library
Abstract Not Provided
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
MCNP6 delayed gamma benchmark using experimental data
Abstract Not Provided
97 MATHEMATICS AND COMPUTING↗