查看更多>>摘要：Grain boundary (GB) character and radiation conditions restrict the in-depth understanding of the effectiveness of tailoring grain size in controlling the radiation damage and promoting the radiation resistance of materials. Here, we constructed 46 [0 0 1] W symmetric tilt GBs with tilt angle varying from 6.73 degrees to 82.37 degrees Then, molecular statics (MS) was used to investigate the impact of GB character (tilt angle, GB energy, E) on the defect energetics. The present results suggested that it is energetically favorable for Vs/SIAs to segregate into the GB and GBs have the biased absorption of SIAs over Vs. Some high-angle and high-energy GBs could provide larger energetic driving force for the V/SIA to segregate. Multiple Vs/SIAs tended to be bound together inside the GB, with a varying binding strength. Using steady-state rate theory (RT) calculations, we explored the Vs accumulation in the W systems with different grain sizes (50-1000 nm) under the typical radiation conditions of temperature (300-1800 K) and dose rate (10(-8)-10(2) dpa/s). In particular, based on the results of MS calculations, the effect of GB character on the Vs accumulation was examined by varying V segregation energies (0.99-2.04 eV). In the process, two extreme values (2.04 and 0.99 eV) of V segregation energy were extracted to investigate its impact, with other values being in this range. The results suggested that nanocrystalline (NC) W would exhibit better radiation tolerance than coarse-grained (CG) W under the conditions of a relatively high temperature and low dose rate. A high V segregation energy led to a low damage accumulation in the grain interior. Some high-angle and high-energy GBs with a large V segregation energy and a small V-V binding energy were expected to improve the radiation resistance of NC. (c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：Fuel fragmentation, relocation, and dispersal are some of the largest issues remaining in the nuclear industry before rod-average burnup can be increased beyond 62 GWd/tU. The issue is primarily related to the potential for fuel to be dispersed into the reactor primary system, which may increase public risk. One way to prevent dispersal is to avoid cladding burst. The objective of this work is to support the high burnup safety case by evaluating cladding burst under high-burnup, full-length fuel rods and to identify uncertainties that could improve model predictions. The results of this analysis will evaluate realistic, prototypic loss-of-coolant accident (LOCA) conditions; support future cladding burst test designs; and inform the development of mechanistic material models. Realistic high-burnup operating conditions were implemented in the BISON fuel performance code to simulate steady-state and LOCA transient fuel rod evolution to the point at which cladding burst occurred. Parametric studies are performed to assess code response to changes in rod internal pressures and heating rates. Results were compared with simulated LOCA experiments to identify inconsistencies between commercial fuel rod analysis and experimental validation. The representative full-length fuel rod LOCA simulation results did not agree with cladding burst tests. Cladding burst tests indicated burst occurring well below (100-150 degrees C) those calculated in the full-length fuel rod LOCA analysis. Further investigation indicated that the cladding burst tests do not appear to be representative for full-length fuel rods. The inconsistency investigated in this work showed that the differences between the BISON simulation and the experiment's cladding burst conditions arise from an incomplete characterization of the cladding surface temperature, detailed rodlet characterization, lack of cladding strain measurements, and uncertainty in the cladding creep and failure models. (c) 2022 Elsevier B.V. All rights reserved.
Wang, QiangJudge, Colin D.Howard, CameronMattucci, Mitchell...
13页
查看更多>>摘要：In this study, we characterized the microstructure evolution and hardness of Inconel X-750 samples that were irradiated in-reactor at several different dose rates and temperatures and at doses up to 84 dpa. The irradiation induced lattice defects, the stability of gamma' precipitates, and the formation of helium bubbles were studied. Detailed statistics regarding the size and density of those features were obtained. The combined effects from irradiation dose, dose rate, and temperature on the disordering and dissolution of precipitates and the formation of bubbles were discussed. The isolated contributions of those microstructural features to the total material strength were calculated based on the Dispersed Barrier Hardening (DBH) model. The spatial distribution of He bubbles, especially the visible large ones ( > 2 nm), were also found to either correlate or anti-correlate with the spatial distribution of solute elements, e.g., Ti, Fe, and Cr, at doses higher than 75 dpa.(C) 2022 Elsevier B.V. All rights reserved.
Machado, Norma Maria PereiraPereira, LuizNeyret, MurielLemaitre, Cecile...
11页
查看更多>>摘要：Borosilicate glasses are generally used as matrices to immobilize nuclear fission products resulting from spent fuel reprocessing. In the high-temperature vitrification process (1200 degrees C), most elements to be contained react chemically with the vitrification additives to form a homogeneous glass melt. Platinum Group Metal (PGM) particles are not incorporated chemically in the melt and therefore are present as suspended particles a few microns in size. These particles exhibit an intense aggregation tendency and consequently the suspensions may present an anomalously high apparent viscosity. These systems are characterized by shear-thinning and thixotropic behaviors. However, the interplay between the rheological behavior and the aggregation degree is poorly understood. In this work, the aggregation mechanisms of a simulated nuclear glass melt containing 3.0 wt.% (1.02 vol.%) of PGM particles were investigated. The impact of the shear stress and time on the PGM aggregation degree was determined using an imposed-stress rheometer at high temperature followed by an imaging analysis procedure via Scanning Electron Microscopy (SEM). We present three different aggregation scenarios and their impact on suspension rheology. Based on the experimental data acquired, a force balance computation was performed to illustrate these three scenarios. (c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：In this paper, the effects of pre-exposure and simulated uneven corrosion "pitting" on liquid metal embrittlement (LME) of T91 ferritic/martensitic (F/M) steel exposed to liquid lead-bismuth eutectic (LBE) at 350 degrees C have been studied by oxygen-specific pre-exposure test, tensile test and ANSYS stress simulation. The results show that a "pit" with a size of 0.8 mm in diameter and 0.4 mm in depth on a cylindrical tensile specimen with a diameter of 3 mm and a gauge length of 15 mm can greatly promote occurrence of LME, regardless of the oxygen concentration dissolved in LBE, leading to the formation of typical quasicleavage on the whole fracture surface. ANSYS stress simulation shows that the presence of such a "pit" can give rise to a strong stress concentration effect and the stress level at the "pit" region before yielding is thus increased by a factor of similar to 1.7, which is responsible for the premature failure of oxide scales and earlier LME crack initiation. The results also alarm that the LME enhancement effect of uneven corrosion should not be ignored during reactor system design especially for thin-walled tubes and under transient conditions, when taking this steel as a candidate material. (C) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：We propose a model describing the high burnup structure inter-granular porosity evolution under irradi-ation. The evolution of the porosity collecting the gas diffusing from the grains is modeled by exploiting a second-order Fokker-Planck expansion of the cluster-dynamics master equations governing the prob-lem, considering nucleation of pores, gas absorption due to the diffusional flow from the grains, size-dependent re-solution of gas from pores due to interaction with fission fragments, vacancy absorption, and pore coalescence. Model predictions on xenon local retention, matrix fuel swelling, and porosity evo-lution are compared to experimental data and to models available in fuel performance codes.(c) 2022 The Authors. Published by Elsevier B.V. This is an open access article under the CC BY license( http://creativecommons.org/licenses/by/4.0/ )
查看更多>>摘要：A large amount of radiocaesium-bearing microparticles (CsMPs) were released into surrounding environment during the Fukushima Daiichi Nuclear Power Plant accident. To clarify the formation mechanism of the Type-A CsMPs, which were released from Units 2/3 and have spherical or ellipsoidal morphologies, oxidation behavior of 304 stainless steel containing 1 wt.% Si at 1200 degrees C in steam atmosphere was investigated in this work. Both Fe2SiO4 and Ni-Fe-Cr phases were formed and distributed in the porous oxide scale. With the progress of oxidation, Fe2SiO4 and Ni-Fe-Cr phases were oxidized, causing severe spallation of the oxide scale somewhere between 90 min and 120 min. Trace amounts of Al and Ti were detected in the silica microparticles transformed from Fe2SiO4 oxidation. Furthermore, thermodynamic calculations were performed with the aid of FactSage software, revealing that: (1) when severe spallation of the oxide scale occurs, high levels of Fe oxides can stably exist in silica-based microparticles distributed in the oxide scale; (2) the Type-A CsMPs may be formed in reducing atmospheres. Based on the experimental and thermodynamic results, a completely new formation mechanism of the Type-A CsMPs is proposed. Silicate matrix is inherited from partially oxidized FeO-bearing silica-based microparticles, which are released from the oxide scale due to spallation. Moreover, diffusion of volatile constituents into FeO-bearing silica-based microparticles occurs in the reactor pressure vessel (RPV), not out of the RPV. This new formation mechanism can well explain many characteristics of the Type-A CsMPs. (C) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：In the present study, it is aimed to be produced glass materials for radiation shielding by adding different proportions (0.02, 0.04, 0.06, 0.08, 0.1, 0.15 and 0.2 mol%) of TeO2 to (79.5-x)SiO2-2.5Al(2)O(3)-5Na(2)O-13B(2)O(3)-xTeO(2) borosilicate glasses (BS). The BS-TeO2 glasses were prepared by melt quenching method using TeO2 and BS powders. The produced glasses' fundamental structural properties were investigated using XRD, UV-vis, and FT-IR spectroscopic techniques. XRD patterns of the all prepared glass samples exhibited amorphous structure. The number of energy transitions in the UV-vis spectrum increased depending on the concentration of TeO2 . Furthermore, produced glasses' physical and mechanical properties were evaluated by calculating the bond strength, packing density, bridged oxygen numbers, and molar oxygen volume. Also, the gamma-ray shielding parameters (mass attenuation coefficient, effective atomic number, electron density, mean free path, half-value layer, and radiation protection efficiency) of the glasses were measured at photon energies in the range of 53.16-661.62 keV using a U-LEGe detector with high resolution. The mass attenuation coefficients of BS-TeO2 glasses were calculated at the same energies by using the MCNP5 code. The obtained values were compared with XCOM data, and a very good agreement was achieved between MCNP5 and XCOM. These results showed that the sample containing 0.2 mol % TeO2 was the best photon attenuation ability among other samples.(C) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：Based on first-principles density functional theory calculations, we investigate the structure, stability and electronic properties of UN/ZrC interface in the context of UN fuel particles dispersed in ZrC matrix, aiming to understand the effects of particle size and interfacial defects on the stability of the dispersion nuclear fuel elements. We first compare different contact configurations of UN/ZrC with (100), (110) and (111) orientations, and identify the most stable structures with the lowest total energy. The calculated binding energies between UN and ZrC in these interfaces are 3.2 similar to 8.5 J/m 2 , indicative of high stability. By using UN/ZrC(100) interface as a prototype probe, we explore UN films of one to nine atomic layers, which reveal that the binding strength oscillates only below three layers and converges quickly with higher thickness. This can be explained by orbital hybridization and charge redistribution that leads to reduced quantum size effect of thin films. We further demonstrate that interfacial defects, including various vacancies and exchanged atoms, generally reduce the binding strength. These results not only shed light on the understanding of UN/ZrC interface at an atomic scale, but also provide valuable guidance for future fabrication and implementation of novel nuclear fuels for practical applications. (c) 2022 Elsevier B.V. All rights reserved.
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