Comparison of neutron dose quantities and instrument and dosemeter readings at representative locations in an MOX fuel fabrication plant

Abstract

The relationships between operational and protection quantities, and values of personal dosemeter and instrument readings have been determined for a recently designed MOX fuel fabrication plant. The relationships between the quantities, and the readings of personal dosemeters are sensitive to both the energy and direction distribution of neutron fluence. The energy distributions were calculated using the Monte Carlo code MCBEND. The direction distribution was addressed by calculating independently, spectral components for which the direction distribution could be reasonably assumed. At representative locations, and for assumed worker orientations, the radiation field is analysed as having, in general, three components—a direct, unidirectional component from the nearest identified discrete source, which is considered incident A-P, several unidirectional components from other such sources which are treated as a rotational component and a scattered isotropic component. The calculated spectra were folded with conversion coefficients for personal dose equivalent, H p(10) slab (A-P, ROT and ISO), effective dose, E, (A-P, ROT and ISO), ambient dose equivalent, H *(10), personal dosemeter (AP, ROT and ISO) and survey instrument response characteristics.