A NOVEL COMPUTATIONAL PLATFORM FOR THE PROPAGATION OF NUCLEAR DATA UNCERTAINTIES THROUGH THE FUEL CYCLE CODE ANICCA

This paper presents the first results of a computational platform dedicated to the propagation of nuclear data covariances, all the way to fuel cycle scenario observables. Such platform, based on in-house codes developed at SCK•CEN in Belgium, both for the creation of the many-randomized nuclear data libraries based on ENDF format and for fuel cycle scenario-studies (known as SANDY and ANICCA, respectively), was employed for the uncertainty assessment of the time-dependent inventory computed from a mono-recycling of Plutonium scenario based on a PWR fleet. An essential part of the procedure that deals with the creation of input data libraries to ANICCA, has been carried out this time by the SERPENT2 code. Due to the fact that its neutron transport and depletion parallelized calculation in 72 cores for up to 1640 days and 60 MWd/kg-HM takes almost one hour, it is feasible to finish a total of 100 ANICCA runs based on randomized input libraries created from ENDF/B-VII.1 neutron-reaction covariances in about one week. Therefore, it is consider that the computation of the output population statistics can be inferred from 100 observables representing time-dependent mass inventories. To mention a few results from the aforementioned NEA/OECD benchmark scenario, it was found out that the relative standard deviation of the accumulated plutonium in the final disposal after 120 years was of 7%, while for curium it corresponded to 8%. Thus, sources of uncertainty arising from neutron-reaction covariances do have an impact in the final quantitative analysis of the fuel cycle output uncertainties.

Measurement of leakage neutron spectra with D-T neutrons and evaluated nuclear data

Benchmarking of evaluated neutron nuclear data libraries was performed for ∼14.8 MeV neutrons on the several targets, such as gallium, graphite, silicon carbide, uranium and tungsten samples. The experiments were performed at China Institute of Atomic Energy (CIAE). The neutron leakage spectra from the samples were measured at 60◦ and 120◦ by a TOF technique with a BC501A scintillation detector. The measured spectra are rather well reproduced by MCNP-4C simulations with the CENDL3.1, JENDL-4.0 and the new release ENDF/B-VIII.0, JEFF-3.3 evaluated nuclear data libraries and so on. There have some difference between experiments and simulations for the elastic and inelastic contributions in the partial energy range. And the discrepancies of the neutron leakage spectra in the MCNP simulations originate simply from the differences in the spectra distributions of the neutron reaction channels in the evaluated nuclear data libraries.

The systematics of neutron reaction cross sections

The parameterized theoretical formulae of excitation functions for (n, 2n) and (n, γ) reactions have been established, and for (n,tot), (n,non), (n,3n), (n,p), (n,d), (n,t), (n,3 He) and (n,α) have been recommended. According to these formulae, the SEF code have been developed for systematics calculation of these reactions. The calculated results with the systematics of the corresponding reactions of the discretional nucleus can be provided by the SEF code in the applied range quickly. At the same time, the comparison of calculated results with experimental and evaluated data can be given graphically.

Neutron reaction studies in the rare earth region: First results for the 162Er(n,2n)161Er physics case

In the present work the first experimental results at near two threshold energies for the 162Er(n,2n)161Er reaction study are presented. The reaction cross section was determined at the energies of 11.0 and 11.3 MeV by means of the activation technique. The experimental method and setup is described.