Nuclear Physics


The company performs criticality safety analyses and evaluations of fissile materials such as processes and fuel assemblies used in nuclear reactors, radiation shielding analyses for spent fuel and other radioactive materials, and core design calculations, using state-of-the-art codes such as MCNP, SCALE and CASMO/SIMULATE.

Criticality safety analyses cover a large range of condition. They are performed for various fuel designs (PWR, BWR, VVER), for wet storage conditions (spent fuel pool) and dry storage/transportation casks, for fresh fuel and spent fuel (utilizing NRC approved burnup credit methodologies), and include both UO2 and MOX fuel. Calculations are performed as part of safety analyses reports, but also to optimize existing and new designs of storage and transportation system from a criticality safety perspective. As an example, the figure below shows the neutron flux distribution from fission processes a cross section of a spent fuel transport cask loaded with content of different reactivity in different cells.