Relationship between nuclear and thermomechanical aspects in fuel rod design

The aim of this paper is to provide an analytical determination of temperature and thermal stress in a cylindrical nuclear fuel rod with cladding, including the actual neutron flux distribution in a unit cell, fur a thermal reactor in steady state. The temperature and stresses are evaluated by solving a set of differential and integral equations, derived from the heat conduction and Hooke's law. Some numerical applications are presented (with reference to a common pressurized water reactor) and discussed. The emphasis is placed on the influence of nuclear parameters on the thermal and mechanical behavior of the fuel element.