Preliminary results on the 233U capture cross section and alpha ratio measured at n_TOF (CERN) with the fission tagging technique

233U is of key importance among the fissile nuclei in the Th-U fuel cycle. A particularity of 233U is its small neutron capture cross-section, which is on average about one order of magnitude lower than the fission cross-section.The accuracy in the measurement of the 233U capture cross-section depends crucially onan efficient capture-fission discrimination, thus a combined set-up of fission and ¿-detectors is needed. A measurement of the 233U capture cross-section and capture-to-fissionratio was performed at the CERN n_TOF facility. The Total Absorption Calorimeter (TAC) of n_TOF was employed as ¿-detector coupled with a novel compact ionization chamber as fission detector. A brief description of the experimental set-up will be given, and essential parts of the analysis procedure as well as the preliminary response of the set-up to capture are presented and discussedPostprint (published version

Fission cross section measurements for 240Pu, 242Pu

This report comprises the deliverable 1.5 of the ANDES project (EURATOM contract FP7-249671) of Task 3 "High accuracy measurements for fission" of Work Package 1 entitled "Measurements for advanced reactor systems". This deliverables provide evidence of a successful completion of the objectives of Task 3.JRC.D.4-Standards for Nuclear Safety, Security and Safeguard

Simulations of a parallel plate Fission Chamber using the MCNPX simulation code

MCNPX in its latest version is able to simulate the transportation of Fission Fragments. It opens the door to Fission Chamber simulations. Such simulations are not straightforward and comparisons with experimental spectra often failed. A procedure is described in the present paper to perform such simulation and to process the result to obtain realistic spectra. Simulated spectra are compared with experimental ones in various different conditions to validate the method and to present its limitation

Mesures de sections efficaces d'actinides mineurs d'intérêt pour la transmutation

Les réacteurs actuels produisent deux types de déchets dont la gestion et le devenir soulèvent des problèmes. Il s agit d abord de certains produits de fission et de noyaux lourds (isotopes de l Américium et du Curium) au-delà de l uranium appelés actinides mineurs. Deux options sont envisagées : le stockage en site géologique profond et/ou l incinération de ces déchets dans un flux de neutrons rapides, c est-à-dire, la transmutation par fission. Ces études font appel à de nombreuses données neutroniques. Malheureusement, les bases de données présentent encore de nombreuses insuffisances pour parvenir à des résultats fiables. L objectif de ce travail est ici d actualiser des données nucléaires et de les compléter. Nous avons ainsi mesuré la section efficace de fission de l 243Am (7370 ans) en référence à la diffusion élastique (n,p) afin de fournir des données indépendantes des mesures existantes dans la gamme des neutrons rapides (1 - 8 MeV). La réaction 243Am(n,f) a été analysée en utilisant un modèle statistique décrivant les voies de désexcitation du noyau composé d 244Am. Ainsi les sections efficaces de capture radiative (n,?) et de diffusion inélastique (n,n ) ont pu être évaluées. La mesure directe des sections efficaces neutroniques d actinides mineurs constitue très souvent un véritable défi compte tenu de la forte activité des actinides mineurs. Pour cela, une méthode indirecte a été développée utilisant les réactions de transfert dans le but d étudier certains isotopes du curium. Les réactions 243Am(3He,d)244Cm, 243Am(3He,t)243Cm et 243Am(3He,alpha)242Am nous ont permis de mesurer les probabilités de fission des noyaux de 243,244Cm et de l 242Am. Les sections efficaces de fission des curiums 242,243Cm(162,9 j, 28,5 ans) et de l américium 241Am sont obtenues en multipliant ces probabilités par les sections efficaces calculées de formation des noyaux composés. Pour chaque mesure, une évaluation précise des erreurs a été réalisée à travers une étude des variances-covariances des résultats présentés. Pour les mesures de la réaction 243Am(n,f), une analyse des corrélations d erreurs a permis d interpréter la portée de ces mesures au sein des mesures existantes.The existing reactors produce two kinds of nuclear waste : the fission products and heavy nuclei beyond uranium called minor actinides (Americium and Curium isotopes). Two options are considered: storage in deep geological site and/or transmutation by fast neutron induced fission. These studies involve many neutron data. Unfortunately, these data bases have still many shortcomings to achieve reliable results. The aim of these measurements is to update nuclear data and complement them. We have measured the fission cross section of 243Am (7370y) in reference to the (n,p) elastic scattering to provide new data in a range of fast neutrons (1 - 8 MeV). A statistical model has been developed to describe the reaction 243Am(n,f). Moreover, the cross sections from the following reactions have been be extracted from these calculations: inelastic scattering 243Am(n,n ) and radiative capture 243Am(n,?) cross sections. The direct measurements of neutron cross sections are often a challenge considering the short half-lives of minor actinides. To overcome this problem, a surrogate method using transfer reactions has been used to study few isotopes of curium. The reactions 243Am(3He, d)244cm, 243Am(3He, t)243cm and 243Am(3He, alpha)242Am allowed to measure the fission probabilities of 243,244Cm and 242Am. The fission cross sections of 242,243Cm(162,9d, 28,5y) and 241Am(431y) have been obtained by multiplying these fission probabilities by the calculated compound nuclear neutron cross section relative to each channel. For each measurement, an accurate assessment of the errors was realized through variance-covariance studies. For measurements of the reaction 243Am(n,f), the analysis of error correlations allowed to interpret the scope of these measures within the existing measurements.BORDEAUX1-Bib.electronique (335229901) / SudocSudocFranceF

Selective activation of failure mechanisms in packaged double-heterostructure light emitting diodes using controlled neutron energy irradiation

Commune au CENBG avec ACENInternational audienc

Measurements of (n,γ\gamma) neutron capture cross-sections with liquid scintillator detectors

A method for measuring (n,$\gamma$) neutron capture cross-sections using liquid scintillator detectors has been investigated. If the response function of the detector is known, and the efficiency as a function of energy is low and approximately constant, then gamma cascades can be counted via a method that is independent of the cascade path provided the detector response is manipulated to allow detection efficiency to be proportional to emitted gamma-ray energy. In this paper, we demonstrate the measurement of efficiency and response functions for a C6D6 liquid scintillator using gamma-ray sources and (p,$\gamma$) reactions on light nuclei. Methods to reproduce the detector response and efficiency data successfully using simulations are presented and discussed. An entire response matrix for the detector has been constructed using a new interpolation technique, allowing weighting functions that force the detector efficiency to be proportional to gamma-ray energy to be calculated. An analysis of the sources of error involved in making (n,$\gamma$) cross-section measurements with this method has been undertaken using Monte-Carlo simulation techniques

Efficiency and response function of a C−6-{6}D6_{6} detector used for (n,γ\gamma) capture reaction cross-section measurements

In this contribution we describe the measurement of the efficiency and response functions of a liquid scintillator dectector (C$-{6}$D$_{6}$) using scaled gamma ray sources and (p, $\gamma$) reactions with light nuclei. The measured functions have also been compared to functions calculated using Monte Carlo simulation techniques. If these two functions are known over a wide range of gamma energy, then gamma cascades generated in a (n,$\gamma$) capture reaction can be counted via a method that is independent of their pathways provide the detector response is weighted by functions that force the detection efficiency to be proportional to the emitted gamma ray energy

Measurement of 242^{242}Pu(n,f) in the [1;2MeV] energy range

International audienceThe design of new generation fast nuclear reactors requires highly accurate cross-section measurements in the MeV energy region. The 242 Pu fission cross section is of particular interest for Pu incineration and nuclear waste production. There are discrepancies around 1 MeV incident neutron energy between libraries and among experimental data. Some data suggest the presence of a strong structure between 1 and 1.2 MeV whereas it is barely visible on some other data and its shape is very different among evaluations. The large majority of the 242 Pu(n,f) measurements have been carried out with respect to the 235 U(n,f) secondary-standard cross section. This introduces a strong correlation between independent measurements and this cross section exhibits structures, in particular a steep increase of +10% at 1 MeV. Therefore, we aim to re-measure the 242 Pu(n,f) cross section relative to the primary-standard 1 H(n,n)p cross section, by using a proton recoil detector. This standard has a very high accuracy (0.4%), is not used for of other 242 Pu measurements, and is structureless. An experiment has been carried out in October 2022 at the MONNET facility in JRC Geel, with incident neutron energies from 0.9 MeV to 2.0 MeV. The experimental setup will be presented, and the analysis procedure will be detailed