The objective of the Core Thermal Model Development program currently in progress at the Battelle Pacific Northwest Laboratory is to develop numerical simulation methods for evaluating thermal hydraulic performance of water-cooled nuclear reactor cores and other components during postulated accident conditions. The primary thrust of this program to date has been continued development of the COBRA computer code as a tool for multidimensional transient simulation applicable to either PWR or BWR cores. COBRA-IV incorporates numerous numerical features which significantly extend computational capabilities beyond that of the widely used COBRA-III-C. One unique feature of COBRA-IV is that it maintains the subchannel connection logic used by COBRA-III-C but applies this logic in a manner similar to the finite-difference methods common in numerical hydrodynamics.