Abstract
An accurate estimation of production of gases due to interactions of neutron in structural materials of nuclear systems is an important aspect within the purview of studies of primary radiation damage. Towards this objective, the indigenous code CRaD, which computes neutron primary radiation damage due to displacements of atoms, has been added with the capability to compute gas production for applications in Indian fast reactors. It includes the computation of gas production cross sections by using standard ENDF-6 procedure, estimation of gases in structural materials due to different types of neutron spectra and propagation of nuclear data uncertainties by Total Monte Carlo (TMC) methodology. A comparison of gas production cross sections from CRaD and NJOY (NJOY-2016.31 and NJOY21) is made and a few discrepancies noted with the processing of charged particle production cross sections using the above versions of NJOY are briefly outlined. A simple linear transmutation methodology is adapted in CRaD to estimate the concentrations of gases and validated with the published theoretical data which were obtained using a similar methodology. The study shows that the representation of charged particle data in different ENDF-6 files is not unique and spectrum-averaged gas production cross sections show large variations. The effect of spread in the activation cross sections from different libraries on the two-step processes of gas production in Nickel under different spectra is also discussed. Detailed TMC-based statistical analyses of gas production in Fe are performed and the mean, standard deviation, skewness and channel-wise and isotope-wise correlation coefficients are quantified. The uncertainties in the production of gases in Fe are estimated to be in the range of 16–56%, with the larger uncertainties being observed in the productions of tritium and helium-3 gases. The gas production parameters have shown non-Gaussian distributions.
Published Version
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