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[en] Uranium dioxide (UO_2) is the most widely used nuclear fuel in existing nuclear reactors around the world. While in service for energy supply, UO_2 is submitted to the neutron flux and undergoes nuclear fission chain reactions, which create large number of fission products and point defects. The study of the behavior of the fission products and point defects is important to understand the fuel properties under irradiation. We conduct electronic structure calculations based on the density functional theory (DFT) to model this radiation damage at the atomic scale. The DFT+U method is used to describe the strong correlation of the 4f electrons of cerium and 5f electrons of uranium in the materials studied (UO_2, CeO_2 and (U, Ce)O_2). (U, Ce)O_2 is studied because it is considered as a low radioactive model material of mixed actinide oxides such as the MOX fuel (U, Pu)O_2 used in light water reactors and fast neutron reactors. Cerium dioxide (CeO_2) is studied to provide reference data of (U, Ce)O_2. We perform a DFT+U study of point defects and gaseous fission products (Xe and Kr) in CeO_2 and compare our results to the existing ones of UO_2. We study the bulk properties as well as the behavior of defects for (U, Ce)O_2, and compare our results to the ones of (U, Pu)O_2. Furthermore, for the study of defects in UO_2, methodological improvements are explored considering the spin-orbit coupling effect and the finite-size effect of the simulation supercell. (author)
[fr]Le dioxyde d'uranium (UO_2) est le combustible nucleaire le plus largement utilise dans les reacteurs nucleaires a travers le monde. En conditions d'exploitation, UO_2 est soumis au flux de neutrons et subit des reactions en chaine de fission nucleaire, ce qui cree un grand nombre de produits de fission et des defauts ponctuels. L'etude du comportement des produits de fission et des defauts ponctuels est importante pour comprendre les proprietes du combustible sous irradiation. Nous effectuons des calculs de structure electronique bases sur la theorie de la fonctionnelle de la densite (DFT) pour modeliser les degats d'irradiation a l'echelle atomique. La methode DFT+U est utilise pour decrire les fortes correlations des electron 4f du cerium et des electrons 5f de l'uranium dans les materiaux etudies (UO_2, CeO_2 et (U, Ce)O_2). (U, Ce)O_2 est etudie car il est considere comme un materiau modele peu radioactif d'oxydes d'actinides mixtes comme (U, Pu)O_2 qui est le combustible d'oxydes mixtes (MOX) utilise dans les reacteurs a eau legere et les reacteurs a neutrons rapides. Le dioxyde de cerium (CeO_2) est etudie pour des donnees de reference de (U, Ce)O_2. Nous effectuons une etude DFT+U des defauts ponctuels et des produits de fission gazeux (Xe et Kr) dans CeO_2 et comparons nos resultats a ceux deja existants pour l'UO_2. Nous etudions les proprietes en volume, ainsi que le comportement des defauts pour (U, Ce)O_2, et comparons nos resultats a ceux de (U, Pu)O_2. En outre, pour l'etude des defauts dans UO_2, des ameliorations methodologiques sont explorees considerant l'effet de couplage spin-orbite et l'effet de taille finie de la supercellule de modelisation
[en] Highlights: • Y_2NiMnO_6 epitaxial films on different substrates were obtained via a PAD method. • All films are ferromagnetic and the Tc of them is lower than the bulk materials. • FM transition temperature Tc decreases with the decreasing the thickness of film. • The magnetic properties of the film are dependent on the type of the substrate. - Abstract: The effects of epitaxial strain induced by lattice mismatch and substrate type on the structure and magnetic properties of Y_2NiMnO_6 thin films have been systematically investigated. Y_2NiMnO_6 thin films grown on (001)-oriented LaAlO_3 (LAO), (La,Sr)(Al,Ta)O_3 (LSAT), and SrTiO_3 (STO) substrates with varying film thickness are obtained by a simple polymer assisted deposition method. X-ray diffraction and Raman scattering observations indicate that the single-phase epitaxial films are successfully obtained. By magnetic measurements, it is found that all the films show an obvious ferromagnetic transition with a lower transition temperatures (Tc) than that of the Y_2NiMnO_6 bulk. With the biaxial tensile strain decreasing or the film thickness increasing, the Tc of the film increases. Besides, due to the different strain states of the films and the different surface migrations of the substrates, the structure and magnetic properties show a strong dependence on the type of substrate. It is suggested that the biaxial tensile strain and substrate type have crucial effects on the structure, magnetic properties and the related Tc of the thin film, which can be utilized to engineer the magnetic properties of the films and the related ferroelectricity.
[en] The thermal response of a film-substrate irradiated by pulsed ion beam is analyzed. In the case of relatively lower ion energy, the irradiating of ion beam could be equivalent to a heat flow conducting into the surface of the film. By using the finite element code, the thermal conduction of titanium films irradiated by pulsed ion beams with an energy of 600 keV, a current of 12 mA, and different beam sizes were simulated. The clear patterns of the thermal conduction were obtained. (author)
[en] The HTR-10 is the first high temperature gas-cooled test module reactor built in China. In the accident analysis, typical design basis accident and beyond design basis accident (BDBA), including the reactivity accident, the loss of external power ATWS, the air ingress accident and water ingress accident into the primary system are selected and detailed analyzed. The results show that the HTR-10 has inherent safety properties. The maximum fuel element temperature will not exceed the limit value 1230 deg. C. The total amount of graphite corrosion maintains no more than 320 kg, the exposure ratio of first coated particles is less than 2.4% in the BDBA. The released radioactivity is limited at a low level and the ability of fuel particles to retain fission products is not corrupted. Even the consequence of the severest hypothetical accident has no safety danger to the reactor. The reactor can shut itself down via its negative temperature coefficient of reactivity in heat-up conditions
[en] The thermal hydraulic calculations of the 10 MW high temperature gas-cooled-test module (HTR-10) are among the most important indications to judge the reactor performance under design conditions. The power distribution, the temperature distribution and the flow distribution of the HTR-10 are calculated for initial and equilibrium core in this paper. The temperature distribution includes the temperature parameters of fuel elements, the helium coolant and the main components in the reactor. In the temperature calculation of fuel elements, several uncertain factors are considered carefully, including non-uniform burnup, power distribution deviation, manufacture deviation of fuel elements, graphite balls mixed with fuel balls in the core, calculation deviation of heat transfer and so on. In the flow distribution calculation, the conservative pebble bed core flow value is selected. The results show that the maximum fuel temperature is much lower than the limitation and the flow distribution can meet the cooling requirement in the reactor core
[en] The author theoretically studied the charge-exchange effects on space charge limited electron and ion current densities of non-relativistic one-dimensional slab ion diode, and compared with those of without charge exchange
[en] Highlights: •Algorithms for burnup calculations with external feed are studied. •Models including relative feed and absolute feed are considered, respectively. •TTA and matrix exponential methods are modified to treat external feed. •The proposed methods are concise and suitable for implementation. -- Abstract: Burnup calculation is an essential task in nuclear reactor physics. Burnup equation is generally used to describe the generation, depletion and decay process in the nuclide system. In some nuclide systems, there may exist physical or chemical exchange with the surroundings, such as the spent fuel reprocessing facilities and some advanced reactors with continuous refueling and discharging like the liquid-metal-fuel reactor. This work studies the burnup calculations with considering external feed. Two external feed models, that is the relative feed and the absolute feed, are addressed. The burnup equations with external feed are solved by utilizing the variation of transmutation trajectory analysis (TTA) method and matrix exponential methods. The proposed models and methods have been implemented in the NUIT (NUclide Inventory Tool) code. Various test cases are calculated for validation, and the numerical results are satisfactory.
[en] Highlights: • The XPZ code is development for the HTR lattice physics computation. • The Dancoff factor obtained from the double-heterogeneous model is used for resonance treatment. • Self-shielding factors are used to account for the spatially self-shielding effect. • The fuel compact is homogenized with preserving the first-collision probabilities. • The double heterogeneity effect is analyzed for the HTR fuel pebble. - Abstract: The XPZ code is developed for the lattice physics computation in the high temperature reactors (HTRs). This paper introduces the methodologies adopted in the XPZ code, including the resonance treatment, equivalent homogenization, transport calculation, and burnup calculation. Emphasis is put on the treatment of the double heterogeneity effect, which involves the following two strategies. Firstly, the Dancoff factor obtained from the double-heterogeneous model is used in the resonance treatment. Secondly, with considering the spatially self-shielding effect in dispersed particles, the fuel compact is equivalently homogenized into a homogeneous media so that the conventional transport calculation can be applied. The effectiveness and accuracy of the methodologies are examined against Monte Carlo solutions. Numerical results demonstrate that the XPZ code is promising for lattice physics computations in HTRs.
[en] Highlights: • Mixture concentration in first-combustion cylinder of direct start is measured. • Factors that affect direct start performances are investigated. • Combustion characteristics of first-combustion cylinder are analyzed. • Hydrocarbon emission is considered to determined control strategies of direct start. - Abstract: This study was conducted to investigate the combustion and emissions characteristics of the first-combustion cylinder in a direct-start process. The explosive energy of the first combustion is important for the success of a direct start, but this combustion was rarely addressed in recent research. For a 2.0 L direct-injection spark-ignition engine, the in-cylinder mixture concentration, cylinder pressure, engine speed and exhaust hydrocarbon concentration were detected to analyze the fuel evaporation, combustion, engine movement and engine emissions, respectively. In the first-combustion cylinder of the direct-start process, the injected fuel was often enriched to ensure that an appropriate mixture concentration was obtained for ignition without misfiring. Approximately one-third of the injected fuel would not participate in the combustion process and would therefore reduce the exhaust hydrocarbon emissions. The start position determined the amount of the total explosive energy in the first-combustion cylinder, and an optimal start position for a direct start was found to be at a 70–80° crank angle before the top dead center to obtain a better combustion performance and lower emissions. A lower coolant temperature increased the maximum explosion energy of the first combustion, but additional hydrocarbon emissions were generated. Because there was almost no problem in the direct-start capability with different coolant temperatures after an idling stop, it was necessary to maintain the coolant temperature when the engine was stopped
[en] Graphite is widely used as moderator, reflector and structural materials in the high temperature gas-cooled reactor pebble-bed modular (HTR-PM). In normal operating conditions or water/air ingress accident, the nuclear graphite in the reactor may be oxidized by air or steam. Oxidation behavior of nuclear graphite IG-110 which is used as the structural materials and reflector of HTR-PM is mainly researched in this paper. To investigate the penetration depth of oxygen in IG-110, this paper developed the one dimensional spherical oxidation model. In the oxidation model, the equations considered graphite porosity variation with the graphite weight loss. The effect of weight loss on the effective diffusion coefficient and the oxidation rate was also considered in this model. Based on this theoretical model, this paper obtained the relative concentration and local weight loss ratio profile in graphite. In addition, the local effective diffusion coefficient and oxidation rate in the graphite were also investigated.