Compound Materials and Nanotechnology

Vol. 381

Vol. 381

Recent Trends in Mass Transport in Solids and Liquids

Vol. 380

Vol. 380

Transport Problems with a Focus on Fluid and Heat Flow

Vol. 379

Vol. 379

Transfer Phenomena in Fluid and Heat Flows IV

Vol. 378

Vol. 378

New Development for Heat Transfer in Solids and Fluid Flow

Vol. 377

Vol. 377

Materials and Technologies in Dentistry: Defects, Processing and Characterization

Vol. 376

Vol. 376

Defects and Diffusion Phenomena in Materials for Nuclear Technologies

Vol. 375

Vol. 375

Transfer Phenomena in Fluid and Heat Flows III

Vol. 374

Vol. 374

Positron Annihilation - ICPA-17

Vol. 373

Vol. 373

Transfer Phenomena in Fluid and Heat Flows II

Vol. 372

Vol. 372

Recent Developments in Mass Transport and Related Phenomena in Materials

Vol. 371

Vol. 371

Transfer Phenomena in Fluid and Heat Flows

Vol. 370

Vol. 370

Diffusion in Solids and Liquids XI

Vol. 369

Vol. 369

# Defects and Diffusion Phenomena in Materials for Nuclear Technologies

375卷

doi: 10.4028/www.scientific.net/DDF.375

文章题目 页数

摘要: Diffusion at infinite dilution of U in metals, with particular emphasis in those used in nuclear facilities, is revisited. Early works present some particularities such as activation enthalpies lower than the vacancy formation enthalpy in the matrix, large differences with self-diffusion in the base material, up to four orders of magnitude differences between measurements performed by different authors in similar temperature ranges, etc. In particular U self-diffusion was qualified as abnormal when compared with other metals. Recent studies by means of α-spectrometry reveal a normal behaviour: activation enthalpies and pre-exponential factors similar to the self-diffusion one and diffusion coefficient values in the same order of magnitude than self-diffusion. The possible influence of short circuits, impurities and/or uncertainties in the techniques used in the early works is discussed in order to explain the differences obtained.

3

摘要: Monolithic fuel system with U – 10 wt.% Mo (U10Mo) fuel alloy has been developed for the Materials Management and Minimization reactor conversion program to replace highly-enriched fuels in research and test reactors with low-enriched fuels. Interdiffusion and phase transformations in the system constituents, i.e., U10Mo fuel, AA6061 cladding, and Zr diffusion barrier, have been investigated using fuel plates fabricated via rolling and hot-isostatic pressing. Diffusion couples, utilizing the constituents of the fuel system were also carried out to help understand the findings from fuel plates based on phase equilibria and diffusion kinetics. Findings from both fuel plates and diffusion couples can provide a comprehensive knowledge to assess, model, and predict the performance of monolithic low-enriched fuel system from fabrication to irradiation. This paper summarizes the experimental results reported from characterization of the fuel plates and diffusion couples with emphasis on interactions at the fuel-cladding, fuel-diffusion barrier, cladding-diffusion barrier, and cladding-cladding interfaces. Constituent phases and relevant diffusion kinetics are compared and contrasted, taking into account differences in thermodynamics and kinetics variables such as pressure, temperature, and cooling rate.

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摘要: Paper presents the results of the growth rate of the interaction layer of uranium-molybdenum dispersed fuel in aluminum matrix and influence of silicon alloying on it. The growth process of amorphous interaction layer depends on the radiation diffusion which is proportional to the fission rate in the power of 1⁄4. The alloying of the matrix by silicon does not lead to a change in the mechanism and kinetics of the interaction layer growth, but only slows it down.

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摘要: UO

_{2}transforms into a superionic conductor at temperatures in excess of 2000 K with oxygen ions becoming mobile and exhibiting collective diffusive dynamics. While the response of UO_{2}to irradiation is of current interest, the possible impact of superionic characteristics on defect dynamics and recovery following radiation has not been explored yet. In the current work, we use atomistic simulations to elucidate the short-time dynamical response of stoichiometric UO_{2}subjected to low energy radiation knocks. We observe that the oxygen ions exhibit a collective behavior that is characterized by frequent hopping across their native lattice sites and forming quasi-one-dimensional string-like structures, which are typical of the superionic state. Approximately, a quarter of the displaced oxygen ions dynamically recover through concerted string-like displacements. Our simulations thus suggest a plausible correlation between defect recovery of irradiated UO_{2}and the characteristic superionic hopping mechanism among the oxygen ions.
43

摘要: The basic properties of PuO

_{2}_{−}*were reviewed, and the equilibrium defects in PuO*_{x}_{2}_{−}*were evaluated from the experimental data of the oxygen potential and electrical conductivity as well as the Ab-initio calculation results. Consistency among various properties was confirmed, and the mechanistic models for thermal property representations were derived.*_{x}
57

摘要: 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.

71

摘要: The oxygen chemical diffusion coefficient in (U, Pu)O

_{2-x}was determined by thermo-gravimetry as functions of the Pu content, oxygen-to-metal ratio and temperature. The surface reaction was considered in the diffusion coefficient determination. The activation energy for the chemical diffusion coefficient was 60 kJ/mol and 65 kJ/mol, respectively, in (U_{0.8}Pu_{0.2})O_{2-x}and (U_{0.7}Pu_{0.3})O_{2-x}.
84

摘要: 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.
91

摘要: Multiscale computational approach is used to evaluate microscopic parameters for description of nitride nuclear fuel. The results of atomistic simulation and thermodynamic modeling allow to estimate diffusivity and concentrations of point defects at various stoichiometric ratios of UN

_{1+}*. The diffusivities of Xe atom were calculated in various equilibrium states. In addition, we obtained the dependence of partial nitrogen pressure on*_{x}*x*and temperature. The results of atomistic simulation were used for modeling of nuclear fuel behavior with use of mechanistic fuel codes.
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