Accident analysis in LMFBR fuel rods by the fenht code

This paper describes the fenht code capability related to the safety requirements in nuclear plants. The code solves the non-linear transient heat-transfer problem for the fuel element of a nuclear reactor, in order to simulate any accidental, operational and emergency power transient with arbitrary initial conditions. The temperature distribution in the fuel, gap and cladding is obtained by a finite-element technique based on minimizing the thermal potential with respect to the temperature vector at the nodes of the finite elements. The non-linear differential matricial equation is linearized by an iterative procedure and solved by the Crank—Nicholson method. Also the thermoelastic stresses in the cladding are valued, by the usual Hooke's law. The code has been applied to the analysis of two reference accidents (incidental power transients) occurring in a liquid-metal fast-breeder reactor (LMFBR); the results are reported and briefly discussed.