Calculated actinide and fission product concentration ratios for gaseous effluent monitoring using Monteburns 3.01

Techniques have been developed at Los Alamos National Laboratory for accura tely calculating certain spent-fuel isotope concentration ratios for pressu rized water reactor assemblies using a linked MCNP/ORIGEN2 code named Monte burns 3.01, without resorting to an assembly or full-core calculation. The effects of various fuel parameters such as the number of radial fuel region s per pin, burnup step size, reactor power, reactivity control mechanisms, and axial profiles have been studied The significance of each factor was de termined A method was also proposed for calculating spent-fuel inventories as a function of burnup for a wide range of reactors and fuel types. It was determined that accurate calculations can be obtained using a three-dimens ional, modified pin cell with seven radial fuel regions and two (flat-flux) axial fuel regions calculated with 2000 MWd/tonne U burnup steps for burnu ps ranging from 0 to 50 000 MWd/tonne U. The calculational technique was be nchmarked to measured values from the Calvert Cliffs Unit I reactor, and go od agreement from the point of view of calibrating a monitoring instrument was found for most cases.