Thermal hydraulic analysis of a nuclear fuel bundle after loss of coolant accident

The paper describes a computer program which traces the temperature transient in the (6x6) nuclear fuel bundle and its channel in post loss of coolant accident period through the following phases: (a) bow-down, (b) period of no convectional cooling and only radiation heat transfer among rods and channel walls, (c) core spray initiation, and (d) reflooding of core. Appropriate models for conduction, convection and radiation heat transfer mechanisms and metal water reactions are used. A finite difference approach is used to solve the differential equations describing the model. A special alogrithm combined with an iteration procedure is used to increase accuracy and reduce computer time. (auth.)