Monte carlo burnup analysis code development and application to an incore thermionic space nuclear power system

In the design of the incore thermionic reactor system developed under the Advanced Thermionic Initiative (ATI), the fuel is highly enriched uranium dioxide and the moderating medium is zirconium hydride. The traditional burn up and fuel depletion analysis codes have been found to be inadequate for these calculations, largely because of the material and geometry modeled and because the neutron spectra assumed for the codes such as LEOPARD and ORIGEN do not even closely fit that for a small, thermal reactor using ZrH as moderator. More sophisticated codes such as the transport lattice type code WIMS often lack some materials, such as ZrH. Thus a new method which could accurately calculate the neutron spectrum and the appropriate reaction rates within the fuel element is needed. The method developed utilizes and interconnects the accuracy of the Monte Carlo Neutron Photon (MCNP) method to calculate reaction rates for the important isotopes, and a time dependent depletion routine to calculate the temporal effects on isotope concentrations. This effort required the modification of MCNP itself to perform the additional task of accomplishing bumup calculations. The modified version called, MCNPBURN, evolved to be a general dual purpose code which can be used for standard calculations as well as for bum-up.