A quantitative estimate on the heat transfer in cylindrical fuel rods to account for flux depression inside fuel

摘要

In a nuclear reactor, the power is limited by thermal rather than by nuclear considerations. The reactor core must be operated at a power level that the temperatures of the fuel and cladding anywhere in the core must not exceed safety limits to avoid damages in the fuel elements. Heat transfer from fuel pins can be calculated analytically by using a flat power density in the fuel pin. In actual practice, the neutron flux distribution inside fuel pins results in a smaller effective distance for the heat to be transported to the coolant. This inherent phenomenon gives rise to a heat transfer benefit in fuel pin temperatures. In this research, a quantitative estimate for transferring heat from cylindrical fuel rods is accomplished by considering a non-uniform neutron flux, which leads to a flux depression factor. This, in turn, alters the temperature inside the fuel pin. A theoretical relationship combining the flux depression factor and a ratio of temperature gradients for uniform and non-uniform is derived, and a computational program, based on finite volume method and energy balance, is developed to validate the considered approximation.