Sensitivity and uncertainty studies for a lead cooled fast reactor core

In this paper, using the publicly available design parameters of the Advanced Lead-cooled Fast Reactor Demonstrator (ALFRED) concept, its reactivity coefficients i.e. the fuel and coolant temperature coefficients and the radial expansion coefficient of the fuel cladding are determined. For this purpose, central difference direct perturbation method using Serpent Monte Carlo transport code and the linear perturbation theory of the transport equation were applied. For the latter, TSUNAMI-3D sequence of the SCALE package was utilized. Additionally, the standard deviation of the effective multiplication factor resulted from the uncertainty of nuclear parameters was quantified. The differently computed fuel temperature coefficients were in relatively good agreement. However, in case of the coolant temperature and fuel cladding expansion coefficients, a large discrepancy was experienced. This originates from disregarding the change in the leakage in case of the linear perturbation theory, which enabled the authors to quantify the influence of the leakage on different reactivity coefficients. From the computed results, it can be concluded that the design of ALFRED satisfies the inherent safety requirements. The uncertainty studies showed that the standard error of the multiplication factor resulted from the uncertainty of the nuclear data is considerable. (author)