Température neutronique

Résumé

vignette|400px|Graphique des fonctions de densité de probabilité de vitesse de la vitesse de quelques gaz nobles à une température de (). Des distributions de vitesse similaires sont obtenues pour des neutrons modérés. La température neutronique, aussi appelée par métonymie « énergie des neutrons », est l'énergie cinétique moyenne d'un neutron libre dans sa population, énergie qui est habituellement donnée en électron-volts (abréviation eV et ses multiples, keV, MeV), la température étant en kelvins (K) ou en degrés Celsius (°C). Cette température est le paramètre, unique, dans la modélisation de la distribution des énergies individuelles des neutrons selon la distribution de Maxwell–Boltzmann, modèle connu pour bien représenter le phénomène d'agitation thermique à l'équilibre thermodynamique. Dans le présent cadre, il est utilisé aussi bien pour les neutrons chauds, thermiques que froids, qui sont modérés (c'est-à-dire ralentis) dans un milieu de même température, au sens thermodyna

Source officielle

https://fr.wikipedia.org/wiki/Temp%C3%A9rature_neutronique

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Preliminary analysis of the PETALE program and comparison to simulations

Thomas Jean-François Ligonnet

This Master thesis focus on a preliminary analysis of the experimental data obtained during the PETALE program in the CROCUS reactor at EPFL. The objective of PETALE is to validate neutron nuclear data for stainless steel and its main elements –namely iron, nickel, and chromium– and to contribute to the reduction of their uncertainties. For this purpose, transmission experiments were performed using activation dosimetry in separate reflectors made of the four materials. The manuscript is separated in three main parts. In the first part, after a presentation of the experimental setup, the measured spectra of the irradiated dosimeters are analysed. The results obtained with the methodology developed at EPFL are compared with those obtained by the collaboration partner CEA. The second part details two supplementary studies. First an experiment was performed and analysed to assess the impact of position uncertainties of core centre dosimeters. It was estimated to be 0.14 % for a position uncertainty of ±5 mm. The second study dealt with the estimation of PETALE’s monitors dead time. A dead time correction model was applied on a stable period experiment, and validated against activation dosimetry experiments. The power underestimation was estimated at 5 % at 80 W. The last part presents a comparison between experimental results and Serpent simulations of PETALE’s experiments using the JEFF-3.3 nuclear data library. In total twenty experiments were modelled and compared with their respective experiment. In this preliminary study, the results show a general agreement in the thermal range. An underestimation of the transparency of nickel and chromium for neutron with an energy of 2 MeV is observed. At higher energies, around 8 MeV, the chromium transparency is strongly overestimated. Finally, only a slight overestimation of the transparency to 2 MeV neutron is observed for iron and stainless steel reflectors. Outlooks are provided for the next steps of the analysis and validation, as well as for prospects on a longer term.

2021

Chargement

Understanding the time-dependent behaviour of a nuclear reactor following an intended or unintended change of the reactor conditions is of crucial importance to the safe operation of nuclear reactors. Safety evaluations of nuclear reactors involve the analysis of normal and abnormal reactor operation states with adequate and validated simulation tools. There is a growing interest in supplementing the results of best-estimate reactor calculations with uncertainties. A major source of uncertainty in reactor calculations is the nuclear data. The propagation of uncertainties allows to study their impact on response values of reactor calculations and to establish safety margins in nuclear reactor safety evaluations. For the simulation of reactor transients, the two-step approach consisting of the generation of group constants in infinite lattice geometry and the full-core calculation using a neutron-kinetic/thermal-hydraulic code system has been state-of-the-art. Nowadays, Monte Carlo codes allow for the determination of group constants based on a three-dimensional full-core model. Thus, effects from neighbouring fuel assemblies can be considered. In this thesis, the applicability of the neutron-kinetic code DYN3D coupled with the thermal-hydraulic system code ATHLET for the simulation of transients in small reactor cores was assessed. Experimental reactivity accident tests carried out at the SPERT III research reactor were simulated. Appropriate group constants were generated with the Monte Carlo code Serpent. The overall good agreement between the calculated and the experimental results prove the applicability of the applied codes for the analysis of such reactor transients. Nuclear data uncertainties were propagated through the simulations of two control rod withdrawal transients in a mini-core model of a pressurised water reactor. The random sampling-based method XSUSA in combination with the SCALE code package was used to determine varied two-group constants. Based on these two-group constants, 1,000 DYN3D-ATHLET calculations were performed and statistically analysed to obtain uncertainty information for the reactor power and the reactivity. Since the assumption of normality could not be maintained for the reactor power, distribution-free Wilks tolerance limits were applied. By sensitivity analyses, the number of delayed neutrons per fission of U-235 and the scattering cross sections of U-238 were identified as the major contributors to the reactor power uncertainty. For a potential improvement of transient simulations, an approach was developed that was comprised of the determination of group constants using a Serpent full-core model at selected times during a transient simulation and the updating of the original infinite lattice group constants with these full-core group constants. Various challenges and possible solutions were identified. The applicability of the group constants obtained by a full-core model was assessed by their application in full-core diffusion calculations.

EPFL2021

Chargement

Diffusion de neutrons de 14 MeV par le carbone 12

Claude Joseph

The angular distributions of the differential cross-sections for scattering of 14,1 MeV neutrons from 12C, Q = 0(0+); – 4,43(2+); – 7,65(0+) and – 9,63(8–) MeV, have been measured using a time-of-flight spectrometer. An indication can be found of the excitation of a wide level at about 10 MeV. A Monte-Carlo programme has been worked up and used to correct the results for multiple scattering; it calculates time-of-flight spectra which can be compared with the measured ones in order to deduce the best values of the cross-sections. The results are finally confronted with other experimental determinations and existing theoretical predictions.

s.n.1967

Chargement

Preliminary analysis of the PETALE program and comparison to simulations

Thomas Jean-François Ligonnet

2021

EPFL2021