Development of the computational software tools to automate the computational analyses of fusion relevant benchmarks

The software tool was developed to automate the validation of the evaluated neutron cross sections files against the benchmarks which provide the differential responses. Specifically it was implemented for the energy and time distributions of neutrons and γ-ray spectra measured with the D-T and $^{252}$Cf(s.f.) neutron sources and simulated by Monte Carlo code MCNP. The master script modifies the MCNP input deck by selecting the desired evaluation, runs MCNP and compares the calculated spectra with measured ones in user defined intervals. The criteria chi-squared, either for intervals or for the whole measured range, was selected to judge about the performance of the evaluated cross section data library. The application of the developed tools for the validation of the ENDF/B-VIII.0, FENDL-3.1d, JEFF.3.3 and JENDL-4.0u libraries against the iron spherical benchmarks with $^{252}$Cf and D-T sources has shown that JEFF-3.3 should be considered as superable over all others libraries for the task of the neutron transport. However all tested libraries underestimate the neutron induced γ- rays leakage from bulk iron by factor of two. The reliability of the validation conclusions was strengthened by inter-comparison of the similar benchmarks but carried out in different labs

Covariances for the 56Fe radiation damage cross sections

The energy-energy and reaction-reaction covariance matrices were calculated for the n + 56Fe damage cross-sections by Total Monte Carlo method using the TENDL-2013 random files. They were represented in the ENDF-6 format and added to the unperturbed evaluation file. The uncertainties for the spectrum averaged radiation quantities in the representative fission, fusion and spallation facilities were first time assessed as 5–25%. Additional 5 to 20% have to be added to the atom displacement rate uncertainties to account for accuracy of the primary defects simulation in materials. The reaction-reaction correlation were shown to be 1% or less

Evaluated displacement and gas production cross-sections for materials irradiated with intermediate energy nucleons

International Conference on Nuclear Data for Science and Technology (ND) -- SEP 11-16, 2016 -- Bruges, BELGIUMWOS: 000426429500034Atomic displacement and gas production cross-sections were obtained for a number of materials to calculate radiation damage and gas production rate in nuclear-and fusion reactors, and neutron spallation sources. An advanced atomistic modelling approach was applied for calculations of the number of stable displacements in materials.Fusion for Energy [F4E-GRT-168.01, F4E-GRT-168.02]The work leading to this publication has been funded partially by Fusion for Energy under the Specific Grant Agreements F4E-GRT-168.01 and F4E-GRT-168.02. This publication reflects the views only of the authors, and Fusion for Energy cannot be held responsible for any use which may be made of the information contained therein

Evaluation of the Neutron Data Standards

With the need for improving existing nuclear data evaluations, (e.g., ENDF/B-VIII.0 and JEFF-3.3 releases) the first step was to evaluate the standards for use in such a library. This new standards evaluation made use of improved experimental data and some developments in the methodology of analysis and evaluation. In addition to the work on the traditional standards, this work produced the extension of some energy ranges and includes new reactions that are called reference cross sections. Since the effort extends beyond the traditional standards, it is called the neutron data standards evaluation. This international effort has produced new evaluations of the following cross section standards: the H(n,n), 6Li(n,t), 10B(n,α), 10B(n,), natC(n,n), Au(n,γ), 235U(n,f) and 238U(n,f). Also in the evaluation process the 238U(n,γ) and 239Pu(n,f) cross sections that are not standards were evaluated. Evaluations were also obtained for data that are not traditional standards: the Maxwellian spectrum averaged cross section for the Au(n,γ) cross section at 30 keV; reference cross sections for prompt γ-ray production in fast neutron-induced reactions; reference cross sections for very high energy fission cross sections; the 252Cf spontaneous fission neutron spectrum and the 235U prompt fission neutron spectrum induced by thermal incident neutrons; and the thermal neutron constants. The data and covariance matrices of the uncertainties were obtained directly from the evaluation procedure