Advanced neutron source reactor fuel element steady-state heat transfer analysis

A steady-state heat transfer code, originally developed for high-flux isotope reactor (HFIR) fuel element analysis, has been converted for use in the advanced neutron source reactor (ANSR) project. The code performs an energy balance along several coolant channels (wide, narrow, and average) and accounts for the effects of oxide growth on the plate surface, plate thermal deformation, and fluid property variations to determine the thermal limits of the core (incipient boiling, critical heat flux (CHF), and maximum fuel centerline temperatures). Spatially and time-varying power density profiles generated by detailed neutronics calculations are used as input along with appropriate engineering uncertainty factors and the core inlet fluid conditions.