Papers by Keyword: Nuclear Fuel

Abstract: Sorption behavior of neodymium (Nd) from cerium (Ce) and samarium (Sm) on anion-exchange resins was investigated by column techniques. The anion exchange studies involved the sorption of Nd, Ce and Sm ions onto Dowex 1-X4 (100–200 mesh) in nitric acid – methanol medium. Separation of neodymium from cerium and samarium was carried out by column method to find out the best separation parameter condition with variating HNO₃ concentration and weight of resin. The best elution profile was found using 0.3M HNO₃ : methanol (1 : 9) as eluent through 1.25 grams of Dowex 1-X4 resin with 97.16 % Sm recovery and 95.59 % Nd recovery, while Ce was retained in the column.

Abstract: Article discusses experimental data on creep of (U,Pu)N and other uranium compounds, and possible mechanism of mass-transfer. Proposed equation describes the following creep features: weak temperature dependence at T < 1000°C, creep acceleration in a fuel with micron-sized grains, and acceleration with the content of second phases formed by impurities and fission products. The difference in creep behavior in reactors with thermal and fast neutrons environmentsis discussed. Comparison of irradiation creep of nitride fuel and properties of cladding materials shows that under parameters of fast reactors and typical design of fuel element it is impossible to implement restraining of external nitride swelling. As initial porosity in the fuel will not compensate the nitride swelling, the cladding of fuel element will work in a mode of following the changing of fuel size. Some suggestions on the cladding material properties are done.

Abstract: Atomistic simulations of radiation impact due to collision cascades in oxide and nitride nuclear fuels are performed in this work using combination of Monte Carlo and molecular dynamics techniques. The key parameters of MFPR code models for the athermal self-diffusivity and irradiation-assisted fission product release from fuel are evaluated. The general solution of Olander's integro-differential equation for the knockout mechanism is developed, which allowed extension of the earlier approaches for the long-lived and stable nuclides.

Abstract: To study the fuel debris treatment at Fukushima Daiichi NPP, information on the behaviour of fuel and structural materials in severely damaged reactors, i.e., oxides and metals of uranium and zirconium is essential. Since sea water was introduced to the reactors, situation of fuel debris became different from that for TMI case. In this paper, phase relations of uranium and zirconium oxides were analyzed by powder XRD method at high temperatures. By the heat-treatment of the mixture of UO₂ and ZrO₂ (U:Zr=1:1) under 10 torr air, UO₂ was oxidized to U₃O₈ over 800 ^oC, The UO₂ like phase appeared again at 1350 ^oC which may be caused by the decomposition of U₃O₈. The oxidation behavior of the UO₂-ZrO₂ system was also investigated by using solid solution sample with different U/Zr ratios under different steam and oxygen pressures. The oxidation of the UO₂-ZrO₂ mixture seemed to be suppressed with decreasing U/Zr ratio. The behavior of fuel materials in the presence of seawater was also discussed as well as that for other structural materials.

Abstract: Porosity-controlled nuclear fuel microsphere is an essential material of fabricating minor actinide-bearing dispersion-type nuclear fuel with the infiltration processes. In this paper, monodisperse and size-controlled spherical oxide nuclear fuel particles with size range of 20μm to 800μm were fabricated by means of microfluidic technology combined with sol-gel process using cerium as a surrogate for plutonium. The porous CeO₂ beads with the density range of 25% to 93% T.D. were successfully prepared by the addition of polyethylene glycol 6000 used as a porogen to the feed broth. The uniform U₃O₈ beads were also prepared at the same experimental conditions as CeO₂ beads prepared, which shows the feasibility of the method for fabricating size-controlled monodisperse nuclear fuel beads.

Abstract: A new type of measurement and recording device for the nuclear fuel transport cask was presented on this paper. With the view to the characteristics of specific application object, the system has been designed with the integration of the MEMS sensors. Several kinds of MEMS sensors include MEMS gyroscopes, MEMS accelerometers, MEMS magnetometers and MEMS pressure sensors are integrated together with embedded CPU, micro GPS receiver, and communication unit. With the conception of the modularization, the multi-subsystem (including: stainless steel sleeve structure, shock absorbers, sealed loader, MEMS sensor components, FLASH memory, power supply components, wireless communication components, etc.) was combined to a new transport cask of nuclear fuel. The low power, low cost, high integrate MEMS sensor system can achieve date of the whole process of nuclear fuel conveying condition, and alarm at the licensing of circumstances beyond. The new nuclear fuel transport cask system include: stainless steel sleeve structure, shock absorbers, sealed loader, MEMS sensor components, FLASH memory, power supply components, wireless communication components. The self-programmed software provides a means to analyze all the data during the whole transport process. The sensors’ output, attitude variety, temperature, acceleration, GPS trace can be analyzed simultaneity or independently. Using the data recorded, System Identification is performed to find the final health status of nuclear fuel.

Abstract: As a result of the conducted calculation experiments, optimum ratio of thickness of coverings to the diameter of the fuel core of the dispersive nuclear fuel was determined. It provides maximum duration of the fuel operating periods. New information about parameters of internal block effect and resonance neutrons capture in a nuclear fuel in the energy range up to 100eV is received. The research has been carried out with the support of the Ministry of Education and Science of the Russian Federation, agreement № 14.B37.21.0473, 2012, August, 3rd.

Abstract: The isothermal sections of the U-Mo-C ternary system have been established at 1000°C and 1400°C, using powder X-ray diffraction, scanning electron microscopy coupled with energy dispersive X-ray analysis for the quantification of U and Mo and differential thermal analysis. The main differences between the two sections are the appearance of liquid phase at about 1230°C, due to the peritectic decomposition of γ-UMo, and the peritectoid decompositions of MoC and β’’ Mo₂C. No other transformation was detected in this temperature range, especially one involving the two only ternary phases found, UMoC₂ and U₂Mo₂C₃.

Abstract: For the last 30 years high uranium density dispersion fuels have been developed in order to accomplish the low enrichment goals of the Reduced Enrichment for Research and Test Reactors (RERTR) Program. Gamma U-Mo alloys, particularly with 7 to 10 wt% Mo, as a fuel phase dispersed in aluminum matrix, have shown good results concerning its performance under irradiation tests. Thats why this fissile phase is considered to be used in the nuclear fuel of the Brazilian Multipurpose Research Reactor (RMB), currently being designed. Powder production from these ductile alloys has been attained by atomization, mechanical (machining, grinding, cryogenic milling) and chemical (hydriding-dehydriding) methods. This work is a part of the efforts presently under way at IPEN to investigate the feasibility of these methods. Results on alloy fabrication by induction melting and γ-stabilization of U-10Mo alloys are presented. Some results on powder production and characterization are also discussed.

Abstract: The Cluster Dynamics method is assessed for the investigation of fission gas behaviour in a krypton-implanted and annealed UO₂ sample. The simulation results are then compared to Thermal Desorption Spectroscopy (TDS) data. A release mechanism is proposed: the initial burst is related to krypton migration via an interstitial mechanism, while the second stage of the release process can be accounted for by the diffusion of krypton in a substitutional position. This latter mechanism is compatible with a diffusion coefficient of 4.10^-21 m²/s.