General Empirical Model for ⁶⁰Co Generation in Pressurized Water Reactors with Continuous Refueling

A simplified model is presented that permits one to calculate the average activity on the fuel elements of a reactor that operates under continuous refueling, based on the assumption of crud interchange between fuel element surface and coolant in the form of particulate material only and using the crud specific activity as an empirical parameter determined in plant. The net activity flux from core to out-of-core components is then calculated in the form of parametric curves depending on crud specific activity and rate of particulate release from fuel surface. In pressure vessel reactors, contribution to out-of-core radionuclide inventory arising in the release of activated materials from core components must be taken into account. The contribution from in situ activation of core components is calculated from the rates of release and the specific activities corresponding to the exposed surface of the component (calculated in a straightforward way on the basis of core geometry and neutron fluxes). The rates of release can be taken from the literature, or in the case of cobalt-rich alloys, can be calculated from experimentally determined cobalt contents of structural components and crud. For pressure vessel reactors operating under continuous refueling, activation of deposited crud and release of activated materials are compared; the latter, in certain cases, may represent a sizable (and even the largest) fraction of the total cobalt activity. It is proposed that the ratio of activities of ⁵⁹Fe to ⁵⁴Mn may be used as a diagnostic tool for in situ activation of structural materials; available data indicate ratios close to unity for pressure tube heavy water reactors (no in situ activation) and ratios around 4 to 10 for pressure vessel, heavy water reactors.