Code and data enhancements of the MURE C++ environment for Monte-Carlo simulation and depletion

Irradiation calculations with MCNP are nowadays an accepted industry standard. Since MCNPX v2.6 followed by MCNP v6 depletion functionality is provided by in-code-coupling of the CINDER code. Depletion statements are placed in the MCNP input and some depletion cycle dependent manipulation options can be used such as total thermal power of the systems, nuclide adaptions to given experimental concentrations etc. However, flexible functionality such as fuel cycle calculations with reshuffling of fuel elements and more complex simulations are not implemented in MCNP at the moment.