62 research outputs found

    Evaluation of the Neutron Data Standards

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    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

    Contributions à l’étude du domaine en énergie des résonances non résolues des réactions induites par neutrons

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    This document presents the statistical description of neutron cross sections in the unresolved resonance range. The modeling of the total cross section and of the "shape - elastic" cross section is based on the "average R-Matrix" formalism. The partial cross sections describing the radiative capture, elastic scattering, inelastic scattering and fission process are calculated using the Hauser-Feshbach formalism with width fluctuation corrections. In the unresolved resonance range, these models depend on the average resonance parameters (neutron strenght function Sc, mean level spacing Dc, average partial reaction widths, channel radius ac, effectif radius R' and distant level parameter). The codes (NJOY, CALENDF...) dedicated to the processing of nuclear data libraries (JEFF, ENDF\B, JENDL, CENDL, BROND... ) use the average parameters to take into account the self-shielding phenomenon for the simulation of the neutron transport in Monte-Carlo (MCNP, TRIPOLI... ) and deterministic (APOLLO, ERANOS...) codes. The evaluation work consists in establishing a consistent set of average parameters as a function of the total angular momentum J of the system and of the orbital moment of the incident neutron l. The work presented in this paper aims to describe the links between the S-Matrix and the "average R-Matrix" formalism for the calculation of Sc, ac and R'.Ce document présente la description statistique des sections efficaces neutroniques dans le domaine en énergie des résonances non résolues. La modélisation de la section efficace totale et de la section efficace "shape-elastic" est basée sur le formalisme de la "Matrice-R moyenne". Les sections efficaces partielles décrivant les réactions de captures radiatives, de diffusion élastique, de diffusion inélastique et de fission sont calculées à l'aide du formalisme Hauser-Feshbach avec fluctuations des largeurs. Dans le domaine des résonances non résolues, ces modèles dépendent des paramètres de résonances moyens ("neutron strenght function" Sc, espacement moyen entre les r\'esonances Dc, largeur moyenne de réaction partielle, rayon de voie ac, rayon effectif R’ et paramètre des niveaux distants). Les codes (NJOY, CALENDF ...) dédiés au traitement des bibliothèques de données nucléaires (JEFF, ENDF/B, JENDL, CENDL, BROND ...) utilisent les paramètres moyens pour prendre en compte le phénomène d'autoprotection des résonances non résolues, indispensable à la simulation du transport des neutrons par les codes stochastiques (MCNP, TRIPOLI ...) et déterministes (APOLLO, ERANOS ...). Le travail d'évaluation consiste à établir un ensemble cohérent de paramètres moyens dépendants du moment angulaire total du système J et du moment orbital du neutron incident l. Les travaux exposés dans ce document s'attachent à décrire les liens entre les formalismes de la Matrice-S et celui de la "Matrice-R moyenne" pour le calcul des paramètres Sc, ac et R’

    Generation of thermal scattering files with the CINEL code

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    The CINEL code dedicated to generate the thermal neutron scattering files in ENDF-6 format for solid crystalline, free gas materials and liquid water is presented. Compared to the LEAPR module of the NJOY code, CINEL is able to calculate the coherent and incoherent elastic scattering cross sections for any solid crystalline materials. Specific material properties such as anharmonicity and texture can be taken into account in CINEL. The calculation of the thermal scattering laws can be accelerated by using graphics processing unit (GPU), which enables to remove the short collision time approximation for large values of momentum transfer. CINEL is able to generate automatically the grids of dimensionless momentum and energy transfers. The Sampling the Velocity of the Target nucleus (SVT) algorithm capable of determining the scattered neutron distributions is implemented in CINEL. The obtained distributions for free target nuclei such as hydrogen and oxygen are in good agreement with analytical results and Monte-Carlo simulations when incident neutron energies are above a few eV. The introduction of the effective temperature and the rejection step to the SVT algorithm shows improvements to the neutron up-scattering treatment of hydrogen bound in liquid water

    Generation of thermal scattering files with the CINEL code

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    International audienceThe CINEL code dedicated to generate the thermal neutron scattering files in ENDF-6 format for solid crystalline, free gas materials and liquid water is presented. Compared to the LEAPR module of the NJOY code, CINEL is able to calculate the coherent and incoherent elastic scattering cross sections for any solid crystalline materials. Specific material properties such as anharmonicity and texture can be taken into account in CINEL. The calculation of the thermal scattering laws can be accelerated by using graphics processing unit (GPU), which enables to remove the short collision time approximation for large values of momentum transfer. CINEL is able to generate automatically the grids of dimensionless momentum and energy transfers. The Sampling the Velocity of the Target nucleus (SVT) algorithm capable of determining the scattered neutron distributions is implemented in CINEL. The obtained distributions for free target nuclei such as hydrogen and oxygen are in good agreement with analytical results and Monte-Carlo simulations when incident neutron energies are above a few eV. The introduction of the effective temperature and the rejection step to the SVT algorithm shows improvements to the neutron up-scattering treatment of hydrogen bound in liquid water
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