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Concept# Neutron cross section

Résumé

In nuclear physics, the concept of a neutron cross section is used to express the likelihood of interaction between an incident neutron and a target nucleus. The neutron cross section σ can be defined as the area in cm2 for which the number of neutron-nuclei reactions taking place is equal to the product of the number of incident neutrons that would pass through the area and the number of target nuclei. In conjunction with the neutron flux, it enables the calculation of the reaction rate, for example to derive the thermal power of a nuclear power plant. The standard unit for measuring the cross section is the barn, which is equal to 10−28 m2 or 10−24 cm2. The larger the neutron cross section, the more likely a neutron will react with the nucleus.
An isotope (or nuclide) can be classified according to its neutron cross section and how it reacts to an incident neutron. Nuclides that tend to absorb a neutron and either decay or keep the neutron in its nucleus are neutron absorbers and will have a capture cross section for that reaction. Isotopes that undergo fission are fissionable fuels and have a corresponding fission cross section. The remaining isotopes will simply scatter the neutron, and have a scatter cross section. Some isotopes, like uranium-238, have nonzero cross sections of all three.
Isotopes which have a large scatter cross section and a low mass are good neutron moderators (see chart below). Nuclides which have a large absorption cross section are neutron poisons if they are neither fissile nor undergo decay. A poison that is purposely inserted into a nuclear reactor for controlling its reactivity in the long term and improve its shutdown margin is called a burnable poison.
The neutron cross section, and therefore the probability of an neutron-nucleus interaction, depends on:
the target type (hydrogen, uranium...),
the type of nuclear reaction (scattering, fission...).
the incident particle energy, also called speed or temperature (thermal, fast...

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Cours associés (19)

Séances de cours associées (192)

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This course is intended to understand the engineering design of nuclear power plants using the basic principles of reactor physics, fluid flow and heat transfer. This course includes the following: Re

Température neutronique

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).

Neutron cross section

In nuclear physics, the concept of a neutron cross section is used to express the likelihood of interaction between an incident neutron and a target nucleus. The neutron cross section σ can be defined as the area in cm2 for which the number of neutron-nuclei reactions taking place is equal to the product of the number of incident neutrons that would pass through the area and the number of target nuclei. In conjunction with the neutron flux, it enables the calculation of the reaction rate, for example to derive the thermal power of a nuclear power plant.

Plutonium 239

Le plutonium 239, noté Pu, est l'isotope du plutonium dont le nombre de masse est égal à 239 : son noyau atomique compte et avec un état fondamental ayant un spin 1/2+ pour une masse atomique de . Il est caractérisé par un excès de masse de et une énergie de liaison nucléaire par nucléon de . Un gramme de présente une radioactivité de , tandis qu'un kilogramme de Pu pur est le siège d'environ par seconde. Il est radioactif avec une période de en produisant de l' par moyennant une énergie de désintégration de .

Interactions des neutrons avec la matièrePHYS-443: Physics of nuclear reactors

Explore l'interaction des neutrons avec la matière, couvrant les sections transversales, les taux de réaction, la diffusion, l'absorption, la dépendance énergétique et l'effet Doppler.

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Couvre l'interaction des neutrons avec la matière, y compris les sections transversales et les mécanismes.

Aluminiures de titane et zirconiumMSE-422: Advanced metallurgy

Discute des propriétés, des applications, de la production et des procédés de séparation des aluminiures de titane, du zirconium et du hafnium.

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An experimental study of long-range magnetic order formation mechanisms in a layered structure with a honeycomb arrangement of the magnetic atoms Na2Ni2TeO6 is conducted. For the first time, the strong spin correlations are directly observed above the Neel temperature T-N that is manifested in the presence of broad diffuse peaks on neutron diffraction patterns obtained with the XYZ polarization analysis. Due to the possibility of separating the magnetic, nuclear incoherent, and nuclear coherent contributions to the total neutron scattering cross section, it is unequivocally established that the observed diffuse scattering has magnetic nature. The spin-pair correlation function is reconstructed by modeling diffuse neutron scattering on Na2Ni2TeO6 with reverse Monte Carlo method. The obtained results indicate 2D nature of the magnetic correlations, and moreover, the symmetry of short-range magnetic state corresponds to long-range zigzag-type magnetic order in the honeycomb net, which is established earlier based on the theoretical calculations.

Andreas Pautz, Jiri Krepel, Fanny Vitullo, Horst-Michael Prasser

In the pebble-bed high-temperature reactor under construction in China, called the HTR-PM, the spherical fuel elements continuously flow downward in the cylindrical core. The burnup of each pebble is checked at the core outlet and, according to the achieved burnup level, the pebble might be disposed or reinserted into the upper section of the core. Upon reinsertion, each pebble is radially distributed in a random manner and, according to its downward path, faces different burnup conditions. Hence, the number of passes necessary to achieve the average discharge burnup of 90 MWd/kgU may vary. Discrete element method (DEM) simulations have been carried out to achieve a clear understanding of the movement of the 420000 fuel pebbles in the HTR-PM core. At the same time, neutronics properties have been investigated for a single pebble and for the full core with the Serpent 2 Monte Carlo code. As a result, one-group microscopic cross sections (XS) have been parametrized at the core level. The pebble movement has been loosely coupled with the depletion of a single pebble in a dedicated burnup script called moving pebble burnup (MPB), developed in MATLAB. 3000 single pebble burnup histories were simulated to obtain sufficient statistics and an insight into the HTR-PM burnup process. The decrease of the average burnup gained per single pass implies that a misshandling of recirculated fuel elements is unlikely to lead to an excess of the maximum allowed burnup of 100 MWd/kgU. The core demonstrates a self-compensation effect of burnup, meaning that it always compensates burnup under- or over-runs in the successive passes. In addition, gamma detection of Cs-137 has been studied as a practical method for monitoring the burnup of the discharged pebbles, turning out to be an applicable measurement technique. Finally, it is possible to conclude that the fuel cycle of the HTR-PM, as it has been laid out, is well designed and feasible.

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.