Abstract
A detailed description is given of a thermohydraulic subchannel analysis of multirod fuel bundles with single phase cooling. Both grid spacers and helical wire spacers are considered. An analytical mathematical approach has been adopted for computing the axial distribution of subchannel coolant temperatures. The computation method has been applied to fast reactor fuel elements with liquid metal cooling. Intersubchannel heat transport mechanisms are discussed in detail. Numerical results centre on the effect a lateral power gradient has on the temperature field in a hexagonal fast reactor fuel rod assembly. An important result obtained, is that for large hexagonal fuel rod assemblies intersubchannel heat transport does not contribute to the attenuation of circumferential flow duct temperature differences. This holds both for assemblies with grid spacers and those with helical wire spacers.
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