The authors provide an overview of the early stages of a new approach for assessing the integrity of AGR graphite cores which incorporates improved methods and accommodates the presence of broken reactor components using a probabilistic treatment. Nuclear Electric has been developing this methodology based upon the probabilistic analysis of core ageing and component failure with a computer code CORSET. The code examines the distortion of the core under various operating conditions, allowing for local discontinuities due to component failure. These distortions are repeatedly assessed allowing for statistical variations in component dimensions and properties, thus constructing a probabilistic distribution against which the required operational and safety features can be confirmed.< >
The energy spectra of neutrons transmitted through iron, lead and perspex slabs were measured for 0·5, 1·0, 1·5 and 1·8 incident neutron energies, using a helium-3 spectrometer. The energy spectra were obtained as a function of the thickness of the scatterer.
Using heterogeneous formalism we calculated Selengut equivalent diffusion coefficient and absorption cross-section for a heterogeneous cell, containing absorbing media. The observed deviation between Benoist and Selengut diffusion coefficients and also between average and Selengut absorption cross-sections are almost exactly compensated by cell-edge flux normalisation in the thermal group.
A continuous flow vanadium bath has been used to measure the neutron yield from different accelerator targets, and the correction factors which have to be applied are discussed. The results of yield measurements for protons from 5 to 10 MeV on thick targets of Li, Be, Cu and Co are given and compared with calculated values. Based on these comparisons comments are made on the accuracy of the cross sections used in the yield calculations.
The angular distributions and energy spectra of keV neutrons scattered from a typical target assembly employing the 7Li (p, n) 7Be reaction have been calculated by the Monte Carlo method for several proton energies, target backing and beam tube wall thicknesses and target radii. A brief discussion is given of possible errors due to scattering in the experimental measurements of neutron activation capture cross-sections, dose equivalents and angular distributions.
