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  • Impact of the cationic homogeneity on Th0.5U0.5O2 densification and chemical durability
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-12-09
    Laurent Claparede, Nicolas Clavier, Adel Mesbah, Florent Tocino, Stéphanie Szenknect, Johann Ravaux, Nicolas Dacheux

    In order to study the effects of cationic homogeneity on the life cycle of Th1-xUxO2 ceramics, including sintering and reprocessing (dissolution) steps, five different ways of preparation were set up, going from the most homogeneous oxalic co-precipitation to a mechanical mixture of the parent oxides. Dilatometric experiments evidenced the better sintering capability for the most homogenous compounds obtained through wet chemistry methods while dry chemistry routes led to poor density values (between 80 and 90 %TD). However, the introduction of an additional mechanical grinding step prior to the powders sintering systematically led to the homogenization of the systems. Improved homogeneity also provide a better chemical durability associated with the congruent dissolution of thorium and uranium in solution during dissolution tests on Th0.5U0.5O2 samples. However, heterogeneous samples led to incongruent behaviors that can be lowered by introducing a grinding step before the sintered samples preparation. Since the impact of the cationic homogeneity must be followed carefully during dissolution, in operando observations of evolving solid/solution interface by ESEM were performed. They allowed imaging the preferential dissolution of uranium-enriched zones and confirmed the significant impact over dissolution rate of the presence of chemical heterogeneities at the interface.

    更新日期:2018-12-10
  • Reductive extraction of lanthanides (Ce,Sm) and its monitoring in LiClKCl/BiLi system
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-12-10
    Wei Han, Zhuyao Li, Mei Li, Wenlong Li, Milin Zhang, Xiaoguang Yang, Yang Sun

    The reductive extraction behavior of Ln (Ce,Sm) and its on-line monitoring were explored in LiClKCl/BiLi system. According to calibration curve, the relationship of peak currents and concentration of Ln (Ce,Sm) measured by square wave voltammetry, the calculated Ln (III) concentration is almost consistent with that analyzed by ICP-AES, and the relative average error is about 5.37%. The effects of Li content in BiLi alloys, temperature and stirring speed on extraction rates, efficiencies and distribution ratios for Ce and Sm were determined, and found that the extraction rates, efficiencies and distribution ratios for Ce and Sm increase with the increasing the temperature, content of Li in BiLi alloys and stirring speed. The extraction rates and efficiencies for Ce(Sm) from LiClKClCe(Sm)Cl3 molten salts are higher than those for Ce(Sm) from LiClKClCeCl3SmCl3 molten salts, and the ones for Ce are much larger than those for Sm due to the formation of Sm (II) checked using XPS and linear sweep voltammetry. The produced Sm (II) could consume some reduced Sm metal: 2 Sm(III) + Sm = 3 Sm(II) , resulting in the decrease of the extraction rate, efficiency and distribution ratios of Sm. The experimental results showed the feasibility of on-line monitoring of reductive extraction process of Ln in LiCL-KCl/BiLi system by electrochemical technique.

    更新日期:2018-12-10
  • First-principles calculations and experiments for Ce4+ effects on structure and chemical stabilities of Zr1-xCexSiO4
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-12-10
    Shuyang Li, Xiaoyong Yang, Jian Liu, Xiaofeng Zhao, Shahid Hussain, Yi Ding, Yong Yi, Tao Duan

    Using Ce atom as the surrogate for tetravalence actinides, the structure stability, chemical bonding and solubility of Zr1-xCexSiO4 have been investigated by combing first principles calculations and static leach test (MCC-1) experimentally to understand the effect of actinides on stability of ZrSiO4. The structural analysis indicates that the high symmetry of [CeO8] polyhedron could well maintain the structural stability of Zr1-xCexSiO4. However, due to the longer and weaker CeO bond than ZrO bond, LRCe is slightly higher than LRZr. Besides, the LRZr and LRCe are highest in acid leaching solution, followed in alkaline leaching solution and deionized water system. Worth noted, the negative solution energy of −22.254 eV indicates the pre-existing Zr vacancy is energetically favorite to accommodate actinides surrogate Ce impurities or quadrivalent actinides with slight structural deformation, showing the strong structural and chemical stabilities of ZrSiO4.

    更新日期:2018-12-10
  • Fabrication of Li4SiO4-Li2ZrO3 composite pebbles using extrusion and spherodization technique with improved crush load and moisture stability
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-12-07
    G. Jaya Rao, R. Mazumder, S. Bhattacharyya, P. Chaudhuri

    The effect of different amount of Li2ZrO3(LZ) (0-15wt%) addition on the properties of composite Li4SiO4 (LS) ceramic pebbles was studied. The Li4SiO4-Li2ZrO3 composite powder was prepared using solid state route at a calcination temperature as low as 900 °C via in-situ method. The composite pebbles were fabricated using a cost-effective and simple technique called extrusion-spherodization. The sintered pebbles were characterized for density, grain size, pore size distribution, crush load and moisture stability. The composite pebbles showed good sintered density (∼87.1%) for LS-5wt% LZ in comparison to LS pebbles (∼80%) when fired at 1000 °C. Moreover, the LS grain size in the composite pebble was significantly reduced by ∼19.7% when compared to LS pebbles. We also found that the average crush load value of the LS-5wt% LZ composite pebbles was improved by nearly 77% (30N) to that of the pure LS pebbles (17N). The LS-5wt% LZ pebbles showed excellent stability to moisture in terms of phase and mechanical crush load retention after 30 days of exposure to humidity. The results achieved in this work were found to be fit to the desired properties for use as a ceramic solid breeder.

    更新日期:2018-12-08
  • Molecular dynamics investigation of grain boundaries and surfaces in U3Si2
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-12-07
    Benjamin Beeler, Michael Baskes, David Andersson, Michael WD. Cooper, Yongfeng Zhang

    Uranium-silicide (U-Si) fuels are being pursued as a possible accident tolerant fuel (ATF). This uranium alloy benefits from higher thermal conductivity and higher fissile density compared to uranium dioxide (UO2). In order to perform engineering scale nuclear fuel performance simulations, the material properties of the fuel must be known. Currently, the experimental data available for U-Si fuels is rather limited. Thus, multi-scale modeling efforts are underway to address this gap in knowledge. Interfaces play a critical role in the microstructural evolution of nuclear fuel under irradiation, acting both as sinks for point defects and as preferential nucleation sites for fission gas bubbles. In this study, a semi-empirical modified Embedded-Atom Method (MEAM) potential is utilized to investigate grain boundaries and free surfaces in U3Si2. The interfacial energy as a function of temperature is investigated for ten symmetric tilt grain boundaries, eight unique free surfaces and voids of radius up to 35 Å. The point defect segregation energy for both U and Si interstitials and vacancies is also determined for two grain boundary orientations. Finally, the entropy change and free energy change for grain boundaries is calculated as a function of temperature. This is the first study into grain boundary properties of U-Si nuclear fuel.

    更新日期:2018-12-07
  • Adsorption mechanisms of cesium at calcium-silicate-hydrate surfaces using molecular dynamics simulations
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-12-05
    J. Bu, R. Gonzalez Teresa, K.G. Brown, F. Sanchez

    Understanding the mechanisms of interactions and immobilization of radionuclides in cement systems at the molecular level is key to developing safe and novel solutions for the long-term storage of nuclear wastes. The molecular level adsorption mechanisms of cesium (Cs+) ions onto tobermorite 9 Å, tobermorite 14 Å, and jennite – three crystalline structural analogues for calcium-silicate-hydrates, the main components in cement-based materials, were investigated using molecular dynamics (MD) simulations. Convergence monitoring of the simulations, using the root mean square displacement with average correlation analysis, indicated that MD trajectories of 10–15 ns were needed given the size of the selected computational cell to properly capture the dynamic process of adsorption. Cs+ ions adsorbed as inner-sphere complexes at the tobermorite surfaces, while they showed lower affinity for jennite. The charge and structure of the surface layer strongly influenced the adsorption mechanisms of Cs+ ions. Strong association with the hexagonal cavities of the silica sites were seen at the tetrahedral SiO4 surface of the tobermorite structures. Cs+ ion adsorption was attributed to co-ion adsorption at the octahedral CaO6 surface of tobermorite 9 Å, cation exchange at that of tobermorite 14 Å, and electrostatic interaction with the surface hydroxyl groups for jennite. Distribution coefficients of Cs+ ions onto cement pastes derived from molecular dynamics by taking the surface structure and dynamics of Cs+ ions directly into account showed range of overlap with experimental results.

    更新日期:2018-12-05
  • Effect of Ti and TiC alloyants on the mechanical properties of W–based armour materials
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-12-05
    E. Tejado, A. Martin, J.Y. Pastor

    The main requirements of tungsten materials for structural divertor applications comprise properties like high thermal conductivity, high-temperature strength and stability, high recrystallization temperature, and enough ductility for an operation period of about two years under massive neutron load [1]. However, the mechanical properties of tungsten commercial products are still one of the main concerns for their use in structural armour applications. With the aim of improving this aspect, two W/Ti based products are presented in this paper: (1) a W-Ti alloy with a Ti solid solution and (2) an UFG microstructure product with TiC dispersed particles; both with the aim of obtaining a suitable fusion armour material with enhanced properties, especially at very high temperatures when pure tungsten suffers strong thermal degradation. It has been reported that strength and recrystallization control can be improved with dispersed TiC particles which inhibits the grain growth. Furthermore, both flexural strength and fracture toughness were twice and even three times higher than the ones observed for our reference pure tungsten produced by the same group and technique, which is, indeed, a great success. However, the intrinsic brittleness of tungsten cannot be enhanced by particle dispersion or solid solution with Ti. On the contrary, intergranular rupture is enhanced even more, and the DBTT is even higher than that of pure W.

    更新日期:2018-12-05
  • Quantification of hardening contribution of G-Phase precipitation and spinodal decomposition in aged duplex stainless steel: APT analysis and micro-hardness measurements
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-12-04
    R. Badyka, G. Monnet, S. Saillet, C. Domain, C. Pareige

    Ageing of cast duplex stainless steels (DSS) is attributed to the decomposition of the ferrite: spinodal decomposition and precipitation of G-phase particles. This leads to an increase in hardness and a decrease in Charpy toughness. According to the literature, spinodal decomposition is accepted to play a major role on the hardening even if the role of G-phase precipitation on mechanical properties is still not clear. This work links microstructural characterization performed using atom probe tomography to micro-hardness of the ferrite for a wide variety of duplex steels (from cast steels with and without Mo to lean steels) aged under different conditions. An attempt to quantify the contribution of both spinodal decomposition and G-phase precipitation is made by applying linear and square superposition principle of Ardell, Orowan and a modified BKS models. The models used are shown to give an excellent estimation of the experimental values of the hardness increase of the ferrite of the cast and lean steels for a wide range of composition and temperature. This work shows that, conversely to what is said in the literature, spinodal decomposition is not systematically the main contributor to hardening.

    更新日期:2018-12-04
  • The effect of irradiation temperature on damage structures in proton-irradiated zirconium alloys
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-12-04
    M. Topping, A. Harte, T. Ungar, C.P. Race, S. Dumbill, P. Frankel, M. Preuss

    A study into the effects of irradiation temperature on the damage structures that form during proton-irradiation has been carried out on two commercial Zr alloys in order to develop a more mechanistic understanding of the effect of niobium on dislocation loop evolution. The two Zr alloys (Zircaloy-2 and Low-Sn ZIRLO™) were proton irradiated to a damage level of ∼2 dpa at 280 °C, 350 °C and 450 °C. Detailed dislocation analysis was carried out using on-axis bright-field scanning transmission electron microscopy combined with spectral imaging and synchrotron x-ray line profile analysis. The analysis revealed a significant difference in the effect of irradiation temperature on loop size between the two alloys. In the case of the Nb-free Zr-alloy (Zircaloy-2), an increase in irradiation temperature results in a marked increase in a-loop diameter, by a factor of ∼7.5 from 280 to 450 °C, and a stark decrease in the dislocation line density. In contrast, the Nb-containing Zr-alloy (Low-Sn ZIRLO™) showed very little variation of loop size and line density over the same radiation temperature range. The STEM-based spectral imaging revealed irradiation-induced nano-clustering found throughout the matrix in Low-Sn ZIRLO™, which is not present in the case of Zircaloy-2. Therefore, it is proposed that Nb plays a crucial role in the evolution of dislocation loops in Zr through the formation of irradiation precipitation throughout the matrix.

    更新日期:2018-12-04
  • Mechanical properties of tungsten irradiated with high-energy protons and spallation neutrons
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-12-03
    Jemila Habainy, Yong Dai, Yongjoong Lee, Srinivasan Iyengar

    In this work, 3-point bending tests and micro-hardness measurements were performed on pure tungsten irradiated with high-energy protons and spallation neutrons in a target of the Swiss Spallation Neutron Source (SINQ). The specimens were irradiated to doses in the range of 1.3–3.5 dpa, with 37–140 appm He, at temperatures between 75 and 110 °C. During 3-point bending tests performed at temperatures up to 500 °C, all the irradiated specimens fractured in the elastic regime at stresses much below those for the unirradiated reference specimens. The micro-hardness measurements indicate significant irradiation induced hardening under all irradiation conditions. Observation of the fracture surfaces of tested irradiated specimens using scanning electron microscopy shows brittle fracture behaviour in general, whereas detailed features depend strongly on irradiation dose and test temperature.

    更新日期:2018-12-03
  • Influence of dual beam ion irradiation and transient heat loading on tungsten surface morphology and erosion
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-12-03
    G. Sinclair, S. Gonderman, J.K. Tripathi, T. Ray, A. Hassanein

    Material testing for plasma facing components (PFCs) in future fusion devices reveals important damage mechanisms that impact lifetime and performance. During operation, PFCs in the divertor region will be subjected to high particle fluxes and heat loads. Tungsten (W) is currently the leading candidate material for divertor PFCs, due to its high melting point, high thermal conductivity, and low tritium retention. Laboratory experiments have effectively characterized mechanisms of damage on W due to different forms of radiation. However, understanding how different species of incident plasma particles interact with one another to affect the resultant strength of the material remains in development. In this study, W samples were exposed to simultaneous ELM-like heat loading and high-flux He+ and D+ ion irradiation at different ELM intensities to further elucidate the complex synergistic effects inherent in a fusion environment. ELM-like heat loading was replicated using a 1064 nm Nd:YAG pulsed millisecond laser at different energy densities. Exposures produced a shale-like microstructure with or without concurrent He+ and D+ ion irradiation. However, adding D+ to a simultaneous laser + He+ irradiation reduced pore formation and inhibited early-stage nanostructure formation. Changes in surface morphology with the addition of D+ could be attributed to super-saturation in the near-surface layer. While the addition of He+ and D+ clearly increased W erosion, the laser energy density did not have as clear of an effect. Increasing the ELM intensity reduced the number of pores, but increased the pore size. Future studies need to explore whether near-surface D impacts pore formation and total He desorption. Continued research on the combined effect of high-flux particle irradiation and transient heat loading will help refine predictions of material performance for ITER and beyond.

    更新日期:2018-12-03
  • Stored energy release in neutron irradiated silicon carbide
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-12-03
    Lance L. Snead, Yutai Katoh, Takaaki Koyanagi, Kurt Terrani

    The purpose of this investigation is to experimentally quantify the stored energy release upon thermal annealing of previously irradiated high-purity silicon carbide (SiC.) Samples of highly-faulted polycrystalline CVD β-SiC and single crystal 6H-SiC were irradiated in a mixed spectrum fission reactor near 60 °C in a fluence range from 5 × 1023 to 2 × 1026 n/m2 (E > 0.1 MeV), or about 0.05–20 dpa, in order to quantify the stored energy release and correlate the release to the observed microscopic swelling, lattice dilation, and microstructure as observed through TEM. Within the fluence of this study the crystalline material was observed to swell to a remarkable extent, achieving 8.13% dilation, and then cross a threshold dose for amorphization at approximately 1 × 1025 n/m2 (E > 0.1 MeV) Once amorphized the material attains an as-amorphized swelling of 11.7% at this irradiation condition. Coincident with the extraordinary swelling obtained for the crystalline SiC, an equally impressive stored energy release of greater than 2500 J/g at the critical threshold for amorphization is inferred. As expected, following amorphization the stored energy in the structure diminishes, measured to be approximately 590 J/g. Generally, the findings of stored energy are consistent with existing theory, though the amount of stored energy given the large observed crystalline strain is remarkable. The overall conclusion of this work finds comparable stored energy in SiC to that of nuclear graphite, and similar to graphite, a stored energy release in excess of its specific heat in some irradiation conditions.

    更新日期:2018-12-03
  • Radiation effect on nano-indentation properties and deformation mechanism of a Ni-based superalloy X-750
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-12-03
    P. Changizian, Z. Yao, C. Lu, F. Long, M.R. Daymond

    Nano-indentation analysis was employed to investigate the mechanical response of heavy ion irradiated X-750 Ni-based superalloy in both solution-treated (ST) and precipitation-hardened (PH) conditions. Helium pre-implantation was carried out at 300 °C up to 5000 appm followed by Ni+-ion irradiation up to 1 dpa at room temperature or 400 °C. Cross-sectional TEM examination was used to characterize the microstructural evolution of the irradiated material and correlate this with nano-scale mechanical test results. Nano-hardness measurements after irradiation at 400 °C showed a similar trend of irradiation-induced hardening, for both ST and PH materials. In contrast, radiation at room temperature resulted in different mechanical responses, with hardening in the ST condition compared to softening in PH. The hardening behaviour was attributed to the creation of irradiation-induced defects including, cavities, Frank loops and small defect clusters; whereas, the γ′-precipitate instability (disordering/dissolution) in PH material resulted in the observed softening. The individual and combined contribution of each type of defects under irradiation hardening were estimated by employing three different obstacle models; results were verified by nano-indentation data for both ST and PH materials. In addition, the softening of the irradiated PH material which results from disordering and dissolution was separately calculated, allowing an estimate of the total yield strength change. TEM analysis of post-indentation microstructure revealed that un-irradiated X-750 deformed by homogenous dislocation motion; however, localized deformation in the form of nano-twins was the dominant deformation mechanism in irradiated X-750.

    更新日期:2018-12-03
  • Microstructure and calorimetric analysis of the U-Mn binary system
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-12-01
    A. Cea, A. Leenaers, S. Van den Berghe, T. Pardoen

    The microstructure, growth kinetics, and the high temperature phase equilibria of the U-Mn binary system are investigated using differential scanning calorimetry. Alloys with various compositions are prepared in a positive-pressure arc melting furnace in order to examine all possible phase changes within the system. Phase composition and morphology were analyzed using XRD, SEM and EDX. Transformation temperatures and enthalpies have been assessed pertaining to: (i) allotropic phase changes αMn → βMn → γMn →δMn, (ii) two eutectic isotherms UMn2 + βMn → L and U6Mn + UMn2 → L and (iii) melting transitions as a function of composition. The development of the two intermetallic phases U6Mn and UMn2 have been observed. A faceted morphology accompanies the formation of UMn2 in a matrix of U6Mn, while grain boundary decomposition is found of UMn2 in αMn. The transformation temperatures have been compared to existing phase diagrams and are slightly higher than reported in literature. Experimental observations of the allotropic transformations of αMn into βMn and of βMn into γMn are in agreement with a calculated phase diagram. The eutectic composition for UMn2 and αMn is found to be 55 wt%Mn, in agreement with previously reported values, while the enthalpy of fusion associated with this eutectic point is found to be 54 kJ mol−1. The enthalpies of formation for the two intermetallic phases U6Mn and UMn2 are equal to be 25.7 kJ mol−1 and 42 kJ mol−1 respectively.

    更新日期:2018-12-01
  • Laser weld geometry and microstructure of cast Uranium-6 wt% niobium alloy
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-12-01
    J.W. Elmer, A.S. Wu, T. DebRoy

    Laser welding was performed on a U6wt%Nb uranium alloy using a 400 W solid state laser with welding speeds from 20 to 2500 mm/s. This speed range spanned melt pool sizes from traditional welding to surface modification and additive manufacturing. With increasing scan speed, the ratio of weld length relative to depth and width increased, with melt pool lengths being more than 5x greater than the width and 10x greater than the depth at the highest speeds. Keyhole mode welds were shown to occur at low speeds, while conduction mode welds occurred at 700 mm/s or higher as the weld depth dropped off more rapidly than width at higher speeds. Microstructures that form at the boundary between the fusion zone and base metal were observed to have a nonconventional appearance consisting of interpenetrating dark and light contrast phases before cells or dendrites appear. Dendrites with secondary arms form from this boundary in keyhole welds and refine to no visible secondary arms near the weld center. Primary and secondary dendrite arms, where present, were shown to refine in size inversely with cooling rate raised to the 0.465 and 0.375 powers respectively. Dendrites were largely absent from the conduction mode welds at higher speeds, and were replaced by a banded microstructure that appears to form by an oscillatory solidification front mechanism.

    更新日期:2018-12-01
  • Irradiation-enhanced precipitation in PH 13-8 Mo maraging steel Corrax
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-12-01
    Ce Zheng, Ryan Schoell, Peter Hosemann, Djamel Kaoumi

    The effects of irradiation on the precipitation behavior of commercial PH 13-8 Mo maraging steel a.k.a. Corrax are investigated through in-situ ion irradiation. Samples of the alloy in its solution annealed state are irradiated up to 10 dpa at 573 and 773 K using 1 MeV Kr ions, in-situ in a Transmission Electron Microscope (TEM) in order to probe irradiation effects on the precipitation usually observed in this alloy under thermal ageing. Indeed, the alloy is known to develop a relatively fine distribution of precipitates during thermal aging which gives the martensitic alloy its strength. The effects of irradiation are substantiated by comparing with the same TEM samples thermally aged at 773 and 873 K for similar amount of experimental time. Both radiation and thermal aging induced segregation and precipitation are characterized using analytical transmission electron microscopy (TEM) techniques. The diffusion coefficients under irradiation are estimated using the point defect balance equations based on Rate Theory and then compared with the thermal diffusion coefficients, demonstrating the accelerated precipitation of β-phase and Laves-phase in the irradiation case at relatively lower temperature is attributed to the radiation-enhanced diffusion. In addition, a numerical model based on classical nucleation and precipitation growth theories for precipitation is introduced and shows a relatively good agreement with the experimental results in terms of precipitate density. This study serves to generate baseline data on ion irradiation effects on Corrax to learn how this steel responds to irradiation.

    更新日期:2018-12-01
  • An approach in the analysis of microstructure of proton irradiated T91 through XRDLPA using synchrotron and laboratory source
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-29
    Argha Dutta, N. Gayathri, P. Mukherjee, Santu Dey, Sudipta Mandal, Tapatee Kundu Roy, Apu Sarkar, S. Neogy, Archna Sagdeo

    The changes in the microstructure of 3.5 MeV proton irradiated T91 Ferritic-Martensitic steel samples with dose has been evaluated using detailed X-ray diffraction line profile analysis of the data collected using both laboratory and the synchrotron source. Different line profile analysis techniques like Williamson-Hall, modified Rietveld method and convolutional multiple whole profile fitting have been applied to evaluate the microstructural parameters such as domain size, microstrain, dislocation density and the character of the dislocation from both the data. The coherent domain size decreases and the corresponding microstrain increases at the first dose of irradiation. With further irradiation, these parameters show a slight recovery and eventually tend towards saturation with increasing dose. The dislocation density follows similar trend as of the microstrain showing saturation at higher doses of irradiation in both the XRD data. However, a systematic change in the character of the dislocation from screw-type to edge-type could be ascertained only from the data obtained using synchrotron source. The Vickers microhardness of the samples is found to increase continuously with increasing irradiation dose supporting the observed change in the dislocation character from screw-type (glissile) to edge-type (sessile). Both EBSD and TEM analysis of the unirradiated and irradiated samples have supported the results obtained from the XRD analysis.

    更新日期:2018-11-30
  • Micron-sized spinel crystals in high level waste glass compositions: Determination of crystal size and crystal fraction
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-30
    C.E. Lonergan, K. Akinloye-Brown, J. Rice, V. Gervasio, N. Canfield, M.J. Schweiger, J.D. Vienna

    The compositions utilized for immobilization of high-level nuclear wastes (HLW) are controlled using glass property models to avoid the deleterious effects of crystallization in the high-level waste (HLW) vitrification melters. The type and size of the crystals that precipitate during melter operations (typically at 1150 °C) and idling (∼1000 °C) are significantly impacted by glass composition and thermal history. This study was conducted to measure the impact of melt composition and heat treatment temperature on crystal size and fraction. A matrix of 31 multi-component glasses canvasing the expected Hanford HLW compositional space was developed and the glasses fabricated and heat treated at 850, 900, and 950 °C. The crystal amounts, as determined by X-ray diffraction, varied from 0.2 to 41.0 wt%. Spinel concentrations ranged from 0.0 to 13.8 wt%. One glass of the matrix did not precipitate spinel and contained 0.2 wt% RuO2, which was assumed to be undissolved from the melting process. All compositions contained crystals in the as-quenched glass. All of the spinel based crystals present in the glasses were less than 10 μm in diameter, as determined by scanning electron microscopy with image analysis. Composition and temperature dependent models were generated using the resulting data and the best model fit was obtained by only considering spinel concentrations (R2 = 0.87). Two glasses were unable to be characterized because of an inability to process the glass under the conditions of this study. Those glasses were utilized to give insight into a potential multi-component constraint to aid in future statistical composition designs.

    更新日期:2018-11-30
  • Fission gas release from UO2 nuclear fuel: A review
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-08-11
    J. Rest, M.W.D. Cooper, J. Spino, J.A. Turnbull, P. Van Uffelen, C.T. Walker

    Gaseous fission product generation, transport, and release can have a large impact on nuclear fuel performance, degrading fuel and fuel–cladding gap properties. Over the past several decades much progress has been made in understanding the key mechanisms of fission gas behavior through investigations with bulk reactor experiments and simplified analytical models. Concurrently, new mechanisms have come to light that can have a strong influence on gas release, especially the unexpected acceleration of fission gas release under high burn-up conditions. Additionally, novel modeling techniques, such as atomistic, mesoscale, and multiscale methods have joined the arsenal of investigative tools. In this review, existing research on the basic mechanisms of fission gas release during normal reactor operation is summarized, and critical areas where further work is needed are identified and discussed.

    更新日期:2018-11-29
  • Radiation induced dissolution of (U, Gd)O2 pellets in aqueous solution – A comparison to standard UO2 pellets
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-28
    Alexandre Barreiro Fidalgo, Mats Jonsson

    The behavior of spent nuclear fuel exposed to groundwater is crucial in the safety assessment of a deep geological repository for spent nuclear fuel. For this reason, leaching experiments on spent nuclear fuel as well as non-radioactive analogues have been conducted for several decades. Although the processes involved can be considered to be fairly well understood, there is a need for further experimental studies whenever new fuel types are introduced. Fuels with burnable absorbers are now in use but very little is known about their behavior under repository conditions. In this work, the impact of burnable absorbers doping (Gd, 3–8%wt.) on the oxidative dissolution of UO2 in an aqueous system was studied in H2O2 and γ-irradiation induced dissolution experiments. The results showed a significant decrease in uranium dissolution and lower reactivity towards H2O2 for (U,Gd)O2 pellets compared to standard UO2. The resulting decrease in the final oxidative dissolution yield was mainly attributed to decreased redox reactivity of the UO2-matrix upon doping. The results of the gamma radiation exposures display an even larger effect of Gd-doping. These findings indicate that other processes are involved in the radiation-induced dissolution of Gd-doped UO2 compared to pure UO2.

    更新日期:2018-11-28
  • Cluster formation and eventual mobility of helium in a tungsten grain boundary
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-27
    C. González, R. Iglesias

    An exhaustive analysis based on density functional theory (DFT) simulations of the accumulation of several He atoms has been performed at the vicinity of a non-coherent W 〈 110 〉 /W 〈 112 〉 interface. The He impurities have been placed both at interstitial positions along the grooves present at each side of the interface and inside the most stable interfacial single vacancy. At such areas, the electronic charge density is lower and the repulsion with the metallic atoms is minimized. Our results show much lower formation energies at both positions studied here as compared to the equivalent bulk cases, confirming the effective great attraction exerted on helium by this kind of interfaces. The most stable groove is completely filled before the system prefers to promote the He atoms to other alternative groove. On the other hand, the vacancy can admit at most seven He atoms, but the successive ones find the best accommodation in the surrounding sites thereafter. This result corroborates the well-known picture of vacancies as efficient sinks for He atoms in W. The binding energy estimation suggests a larger attraction between the He atoms and the vacancy. From the low values obtained at the interface and the energy barriers estimated, we can infer a decreasing mobility of the He clusters along the interface for a given temperature. This situation could favor their accumulation in the stable grooves until they are filled and the outgassing process could subsequently take place. Therefore, a tungsten system with many interfaces, the so-called nanostructured W, can be considered as a good candidate for plasma facing material in a future nuclear fusion reactor.

    更新日期:2018-11-28
  • The role of chemical disorder and structural freedom in radiation-induced amorphization of silicon carbide deduced from electron spectroscopy and ab initio simulations
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-28
    Alexander J. Leide, Linn W. Hobbs, Ziqiang Wang, Di Chen, Lin Shao, Ju Li

    Chemical disorder has previously been proposed as an explanation for the anomalously facile amorphization of silicon carbide (SiC), on the basis of topological connectivity arguments alone. In this exploratory study, “amorphous” (formally, aperiodic) SiC structures produced in ab initio molecular dynamics simulations were assessed for their connectivity topology and used to compute synthetic electron energy-loss spectra (EELS) using the ab initio real-space multiple scattering code FEFF. The synthesized spectra were compared to experimental EELS spectra collected from an ion-amorphized SiC specimen. A threshold level of chemical disorder χ (expressed as the ratio of the number of carbon-carbon bonds to the number of carbon-silicon bonds) was found to be χ ≈ 0.38, above which structural relaxation resulted in formally aperiodic structures. Different disordering methodologies resulted in identifiably different aperiodic structures, as assessed by local-cluster analysis and confirmed by collecting near-edge electron energy-loss spectra (ELNES). Such structural differences are predicted to arise for SiC crystals amorphized by irradiations involving different damage mechanisms—and therefore differing disordering mechanisms—for example, when contrasting the respective amorphized products of ion irradiation, neutron irradiation, and high-energy electron irradiation. Evidence for sp2-hybridized carbon bonding is observed, both experimentally in the irradiated sample and in simulations, and related to connectivity topology-based models for the amorphization of silicon carbide. New information about the probable intermediate-range structures present in amorphized silicon carbide is deduced from enumeration of primitive rings and evolution of local cluster configurations during the ab initio-modelled amorphization sequences.

    更新日期:2018-11-28
  • High-temperature stability of (Gd1−xCex)2Zr2O7+x (x = 0–1) synthesized by spray pyrolysis in air
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-22
    Xin Wang, Min Xue, Songbai Liu, Kuo Jiang

    In this paper, (Gd1−xCex)2Zr2O7+x immobilized matrices with high cerium (Ce) amount-of-substance fractions (up to 90%) were prepared by sol-spray pyrolysis, and then sintered at 1600 °C in air. XRD results showed that all samples were in a mixture of pyrochlore and fluorite phase, even after sintering at 1600 °C. All the sintered bulks were dense, and the densities were independent of the additive content. The SEM results showed that all samples were chemically homogeneous. There was no segregation in the grain boundary. EDS quantification yielded a mean composition close to the stoichiometric chemical composition.

    更新日期:2018-11-24
  • Mechanisms of radiation-induced segregation around He bubbles in a Fe-Cr-Ni crystal
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-22
    B. Kombaiah, P.D. Edmondson, Y. Wang, L.A. Boatner, Y. Zhang

    Fe-15Cr-15Ni (wt.%) austenitic stainless steel was irradiated with 200 keV 4He+ ions to a fluence of 5 × 1016/cm2 at 500 °C. Radiation-induced segregation (RIS) at He bubbles in the irradiated steel was measured using Energy Dispersive X-ray Spectroscopy (EDS) coupled with Scanning Transmission Electron Microscopy (STEM). The RIS measurements exhibited an enrichment of Ni and a depletion of Cr and Fe at the bubbles. To determine the underlying mechanisms of the RIS, the prediction of the Marwick model was compared with the RIS measurements. The model with parametric input in accordance with the Inverse Kirkendall effect (IKE), i.e. a preferential flux of elements associated with vacancy flux, was inadequate in explaining the observed amount of segregation of Fe, Cr and Ni. Specifically, a considerably lower ratio of the concentration gradient between Cr and Fe, ∇Cr∇Fe, of 0.53 was obtained from the RIS measurements versus the model-predicted value of 2.3, indicating an overestimation of Cr depletion at the bubbles by the Marwick model. To explain this difference between the model prediction and the RIS measurements, in addition to the Cr depletion at the bubbles by IKE due to the fastest diffusion via vacancy flux, a preferential interstitial flux of Cr from the matrix to the bubbles is discussed as a possible concurrently operating mechanism.

    更新日期:2018-11-24
  • Thermal conductivity of plasma deposited amorphous hydrogenated boron and carbon rich thin films
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-23
    Hari Harikrishna, William A. Lanford, Sean W. King, Scott T. Huxtable

    Owing to a unique combination of low density, high mechanical stiffness, and high neutron capture cross section, boron rich solids are of interest for a number of applications ranging from fusion shielding and neutron absorbing materials to solid state neutron detection. In many of these applications, the boron containing materials may be exposed to extreme temperatures and thermal gradients where heat dissipation is a significant concern. While there have been several reports of the thermal conductivity for crystalline boron solids, comparitively little is known regarding the thermal conductivity for amorphous boron thin films used in many nuclear applications. In this investigation, we report time-domain thermoreflectance (TDTR) measurements for a series of plasma deposited boron rich thin films with nominal compositions of a-B:H and a-BP:H. The results were compared to additional measurements performed on a-C:H films with similar mass density and which are also utilized as fusion plasma shielding materials. The values of thermal conductivity determined by TDTR for the a-B:H and a-BP:H films ranged from 1 to 2 W/mK. These values are reduced relative to those reported for crystalline boron (4–400 W/mK) but compare well to those obtained for a-C:H (0.5–1.5 W/mK).

    更新日期:2018-11-24
  • Evolution of radiation-induced lattice defects in 20/25 Nb-stabilised austenitic stainless steel during in-situ proton irradiation
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-19
    C. Barcellini, R.W. Harrison, S. Dumbill, S.E. Donnelly, E. Jimenez-Melero

    We have monitored in situ the lattice defect evolution induced by proton irradiation in 20Cr-25Ni Nb-stabilised stainless steel, used as fuel cladding material in advanced gas-cooled reactors. At 420 °C, the damaged microstructure is mainly characterised by black spots and faulted a 0 3 〈 111 〉 Frank loops. Defect saturation is reached at only 0.1dpa. In contrast, at 460 °C and 500 °C proton bombardment induces the formation of a mixture of a 0 3 〈 111 〉 Frank loops and perfect a 0 2 〈 110 〉 loops. These perfect loops evolve into dislocation lines that form a dense network. This transition coincides with the saturation in the dislocation loop size and number density at 0.8dpa (460 °C) and 0.2dpa (500 °C), respectively. The presence of a high density of dislocation loops and lines at those two temperatures causes a vacancy supersaturation in the matrix, leading to the formation of voids and stacking fault tetrahedra.

    更新日期:2018-11-20
  • Multiscale modeling of crystal plasticity in Reactor Pressure Vessel steels: Prediction of irradiation hardening
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-20
    Ghiath Monnet, Ludovic Vincent, Lionel Gélébart

    The plastic behavior of irradiated Reactor Pressure Vessel (RPV) steels is described by constitutive equations capturing the temperature and strain rate sensitivities. The flow stress is decomposed into its fundamental components associated with the microstructure features peculiar to RPV steels, such as carbides, dislocation network and deformation confinement inside grains. Dislocations are assumed to move on the {110} and {112} crystallographic planes and a simplified interaction matrix is proposed. The predicted yield stress is obtained without adjustable parameters and found in close agreement with a large number of experimental results over a large temperature range. Finally, the contribution of radiation defects is accounted for using atomistic and dislocation dynamics results. The effect of solute cluster is analyzed in details in terms of the cluster size and density and strength. Results are discussed and compared with an experimental database on neutron-irradiated RPV steels.

    更新日期:2018-11-20
  • Cold sintering and durability of iodate-substituted calcium hydroxyapatite (IO-HAp) for the immobilization of radioiodine
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-16
    Muhmood ul Hassan, Ho Jin Ryu

    Cold sintering of iodine-hosted calcium hydroxyapatite was investigated for the development of a durable matrix for radioiodine immobilization. Single-phase, nano-crystalline, iodate-substituted calcium hydroxyapatite (IO-HAp) was synthesized by a wet precipitation method and sintering of the dried IO-HAp powder containing ∼7wt.% of substituted iodine was carried out at 200 °C under a uniaxial pressure of 500 MPa. It was demonstrated that a sintered relative density of 96.8% can be achieved without affecting the iodate nature of the substituted iodine. A product consistency test of the sintered samples was also carried out under a standard condition. The normalized leaching rates of Ca, P and I after seven days were 6.9 (±0.5) × 10−7, 2.6 (±0.2) × 10−7 and 2.4 (±0.4) × 10−5 g/m2/d, respectively, providing evidence of the durability of the cold sintered matrix and promise of using the cold sintering process for the conditioning of volatile element-bearing radioactive waste into solid waste forms.

    更新日期:2018-11-16
  • Analysis of deposited layers with deuterium and impurity elements on samples from the divertor of JET with ITER-like wall
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-16
    P. Ström, P. Petersson, M. Rubel, E. Fortuna-Zaleśna, A. Widdowson, G. Sergienko,

    Inconel-600 blocks and stainless steel covers for quartz microbalance crystals from remote corners in the JET-ILW divertor were studied with time-of-flight elastic recoil detection analysis and nuclear reaction analysis to obtain information about the areal densities and depth profiles of elements present in deposited material layers. Surface morphology and the composition of dust particles were examined with scanning electron microscopy and energy-dispersive X-ray spectroscopy. The analysed components were present in JET during three ITER-like wall campaigns between 2010 and 2017. Deposited layers had a stratified structure, primarily made up of beryllium, carbon and oxygen with varying atomic fractions of deuterium, up to more than 20%. The range of carbon transport from the ribs of the divertor carrier was limited to a few centimeters, and carbon/deuterium co-deposition was indicated on the Inconel blocks. High atomic fractions of deuterium were also found in almost carbon-free layers on the quartz microbalance covers. Layer thicknesses up to more than 1 μm were indicated, but typical values were on the order of a few hundred nanometers. Chromium, iron and nickel fractions were less than or around 1% at layer surfaces while increasing close to the layer-substrate interface. The tungsten fraction depended on the proximity of the plasma strike point to the divertor corners. Particles of tungsten, molybdenum and copper with sizes less than or around 1 μm were found. Nitrogen, argon and neon were present after plasma edge cooling and disruption mitigation. Oxygen-18 was found on component surfaces after injection, indicating in-vessel oxidation. Compensation of elastic recoil detection data for detection efficiency and ion-induced release of deuterium during the measurement gave quantitative agreement with nuclear reaction analysis, which strengthens the validity of the results.

    更新日期:2018-11-16
  • Adsorption and immobilization of radioactive ionic-corrosion-products using magnetic hydroxyapatite and cold-sintering for nuclear waste management applications
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-16
    Suriya Venkatesan, Muhmood ul Hassan, Ho Jin Ryu

    A simple and efficient method for the adsorption and immobilization of the radioactive ionic-corrosion-products Co2+, Cr3+, Mn2+, Fe2+, Ni2+, Cu2+ and Zn2+ generated in nuclear reactor coolant has been developed using a magnetic hydroxyapatite nanocomposite (MA-HAP) and a cold-sintering technique. The structure, morphology, magnetic properties and zeta potential of synthesized MA-HAP adsorbent were studied to evaluate its suitability for cationic uptake. The incorporated magnetic nanoparticles significantly improved the Co2+ adsorption capacity of MA-HAP to 68.95 mg/g under optimized conditions such as a pH of 6, a contact time of 120 min and an adsorbent dosage of 4 g/L. More than 92% of Co2+ was removed from the simulated aqueous solution in the presence of other ionic-corrosion-products. Densification of the nanocomposite loaded with the corrosion product was carried out using a pressure-assisted cold-sintering technique at 200 °C for 10 min. The sintered waste form showed high relative density (>95%) and hardness (>2.5 GPa). The results of a product consistency test indicated a low normalized leaching rate in the range 10−6 to 10−7 g/m2/day for all ions adsorbed by the MA-HAP. Thus, the material and methods introduced here are highly capable of adsorbing and immobilizing radioactive waste.

    更新日期:2018-11-16
  • Failure behavior of SiC/SiC composite tubes under strain rates similar to the pellet-cladding mechanical interaction phase of reactivity-initiated accidents
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-15
    M. Nedim Cinbiz, Takaaki Koyanagi, Gyanender Singh, Yutai Katoh, Kurt A. Terrani, Nicholas R. Brown

    The mechanical response of a nuclear-grade silicon carbide fiber-reinforced silicon carbide matrix (SiC/SiC) composite was investigated under mechanical loading conditions mimicking the pellet-cladding mechanical interaction (PCMI) phase of a reactivity-initiated accident (RIA). In a RIA, cladding deformation and failure can be induced by the rapid thermal expansion of the nuclear fuel. A pulse-controlled modified-burst test was used to investigate RIA-like PCMI scenarios on SiC/SiC composite samples at pulse widths from 12 to 100 ms. The strain-driven nature of the cladding sample deformation was due to the rapid internal pressurization and subsequent expansion of a secondary tube. A digital-image correlation technique was used to measure strains from the speckle-painted outer surface of the tubes. The failure strains of samples tested at slower rates, such as RIA event durations of 52 and 100 ms, showed good agreement with the literature-reported values for similar composites tested at slow strain rates. Additionally, the failure strain showed good agreement with reference expansion-due-to-compression tests at slow strain rate. However, a decrease in the failure strain was determined for the fast-rate (12 ms) tests. This indicated that the failure strain of these composites might be influenced by the strain rate during RIA-like events. The failure strains observed in the tests corresponded to local energy depositions of approximately 50 cal/g UO2 from hot zero power, with an initial condition of pellet–cladding gap closure prior to the event. In-pile transient testing of these concepts that would result in hoop strain due to PCMI in the range of 0.5–1.0% is recommended.

    更新日期:2018-11-15
  • Effect of ion and neutron irradiation on oxide of PHWR fuel tube material
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-15
    Gargi Choudhuri, P. Misra, S. Basu, N. Gayathri, P. Mukherjee, V. Kain, D. Mukherjee, D. Srivastava, G.K. Dey

    The effect of heavy ion irradiation on the stability of oxide phase of autoclaved Zircaloy 4 fuel tube material, has been studied using Glancing angle X-ray diffraction (GIXRD) technique after 306 KeV Ar+9 ions irradiation at a dose of 3 × 1019 Ar+9/m2. To estimate the extent of damage, a simulation was carried out using “Stopping and Range of Ions in Matter (SRIM-2008)” computer program based on the Monte Carlo method. For the first time, the oxide formed in 0n1 irradiated fuel tube after 7600 MWD/T burn up in Pressurized Heavy Water Reactor (PHWR) has been characterized using GIXRD technique. The advantage of this technique is that the stress-induced phase transformation, which normally occurs during metallographic sample preparation for optical and electron microscopy, is eliminated. The un-irradiated autoclaved oxide in the steam environment (415 °C and 500 °C), both uniform as well as nodular oxide has been characterized using GIXRD, X-ray photo-electron spectroscopy (XPS), scanning electron microscopy (SEM) attached with Energy Dispersive spectroscopy (EDS). Cross-sectional transmission electron microscopy has been carried out in uniform oxide before and after heavy ion irradiation. It is observed that heavy ion and neutron irradiation induce monoclinic to tetragonal phase transformation in the oxide. Presence of significant fraction of tetragonal ZrO2 phase as well as sub-oxide Zr3O has been identified in the oxide layer near oxide-coolant interface in 0n1 irradiated in-pile sample whereas in unirradiated autoclaved oxide these phases are present in a small fraction near the oxide-metal interface. XPS analysis indicates the difference in the chemical state of alloying element in the oxide when autoclaving is carried out at different temperatures.

    更新日期:2018-11-15
  • A study on the corrosion and stress corrosion cracking susceptibility of 310-ODS steel in supercritical water
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-14
    Zhao Shen, Kai Chen, Xianglong Guo, Lefu Zhang

    General corrosion and stress corrosion cracking (SCC) susceptibility of an oxide dispersion strengthened (ODS) austenitic 310 (310-ODS) steel in supercritical water (SCW) are studied by weight gain and slow strain rate tensile (SSRT) testing, respectively. 310-ODS steel shows an excellent general corrosion resistance in SCW at 600 °C. Energy dispersive X-ray (EDX) and electron backscattered diffraction (EBSD) are conducted on the cross-section of the surface oxide film, revealing a double-layered structure. Results from SSRT tests at 600 °C show an intergranular fracture mode, and SCC susceptibility of 310-ODS steel increases with the increasing of dissolved oxygen (DO) concentration. SSRT tests at 650 °C show ductile fracture mode, and SCC susceptibility is minimum. ODS enhances the yield strength of 310-ODS steel to over 480 MPa, tensile strength to over 750 MPa, and still keeps elongation rate to over 12% at 600 °C. Combining with its low general corrosion rate and low SCC susceptibility, 310-ODS steel is supposed to be a promising material for the fuel cladding of supercritical water-cooled reactor (SCWR).

    更新日期:2018-11-14
  • Characterization of surface layers formed on DU10Mo ingots after processing steps and high humidity exposure
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-14
    Tiffany C. Kaspar, Christina L. Arendt, Derek L. Neal, Shawn L. Riechers, Crystal Rutherford, Alan Schemer-Kohrn, Steven R. Spurgeon, Lucas E. Sweet, Vineet V. Joshi, Curt A. Lavender, Rick W. Shimskey

    The design of monolithic UMo fuel elements fabricated from low-enriched uranium for use in high-power research reactors requires bonding of the fuel foil to either Al cladding or a Zr barrier layer. Processing of the UMo ingot to final foil form has the potential to generate surface layers on the foil that differ from the bulk, metallic UMo. The interfacial properties between the UMo and Zr or Al cladding layers will then be determined by these surface layers. We use x-ray photoelectron spectroscopy, cross-sectional scanning electron microscopy, and atomic force microscopy to characterize the composition, oxidation state, and morphology of the surface layers that form after hot rolling and cold rolling depleted U–10 wt% Mo alloy (DU10Mo). A thick uranium nitride layer is observed after hot rolling, although its origin is likely from a previous processing step. The efficacy of acid etching in HNO3 is compared to that of electropolishing in H2SO4 to remove surface nitride and oxide layers, and both methods are found to be similarly effective. Both laboratory (low humidity) air exposure and longer rinse times in water are shown to promote the formation of surface oxide layers. Exposure of both acid-etched and electropolished DU10Mo foils to humid air (97% relative humidity) for six weeks results in formation of a thick oxide layer due to corrosion. The oxide layer on the acid-etched foil is thicker and more highly oxidized than the oxide layer that forms on the electropolished foil, and these differences in oxidation behavior are attributed to higher surface roughness on the acid-etched foil. In general, Mo is found to play a role as a sacrificial element, typically exhibiting a larger ratio of Mo6+/Mo4+ than U6+/U4+. This is unexpected, given the greater thermodynamic driving force to form U oxides than Mo oxides.

    更新日期:2018-11-14
  • Intermetallic Re phases formed in ion irradiated WRe alloy
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-14
    R.W. Harrison, G. Greaves, J.A. Hinks, S.E. Donnelly

    Intermetallic Re precipitation at concentrations below the solubility limit is a puzzling phenomenon in neutron irradiated W. Ion irradiation has been unable to reproduce their formation, denying the community the ability to accurately simulate neutron damage microstructures and probe precipitate formation. We have recently been successful in inducing σ (WRe) and χ (WRe3) phase formations in W26Re irradiated with 350 keV Ne ions at 500 and 800 °C. The precipitation of these phases is related to the effects of cascade energy density and ballistic mixing during previous high energy self-ion irradiations is concluded to have caused redissolution of precipitates and thus prevented their observation.

    更新日期:2018-11-14
  • Proton irradiation for radiation-induced changes in microstructures and mechanical properties of austenitic stainless steel
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-14
    Hyung-Ha Jin, Seong Sik Hwang, Min Jae Choi, Gyeong-Geun Lee, Junhyun Kwon

    We report on microstructural and mechanical property changes as a function of radiation damage value in proton-irradiated austenitic stainless steel by means of advanced characterization techniques. The microstructural changes in proton-irradiated austenitic stainless steel were analyzed by transmission electron microscopy for observation of radiation-induced defects as well as the measurement of the chemical composition at grain boundaries. The radiation hardening after the proton irradiation was characterized by nano indentation for changes in hardness profiles with radiation damage.Various transition points for microstructural and mechanical property changes under proton irradiation are analyzed via material characterization of proton-irradiated austenitic stainless steels. The saturation is expected to occur at approximately 10 displacements per atom (dpa) for the radiation-induced segregation of Cr, Ni, and P and approximately 2.5 dpa for radiation hardening. The cavity formation is observed to occur at hydrogen concentration levels greater than 5E5 atomic parts per million (appm) H. It is also found that the transition from black dot to Frank loop happened above approximately 1 dpa.Profiles of radiation-induced segregation and radiation hardening as a function of dpa can be extended to the high irradiation condition, and can be compared with experimental data for neutron irradiation-induced segregation and radiation hardening. The radiation-induced segregation after the proton irradiation at 360 °C are in good agreement with that after neutron irradiation. On the other hand, it is observed that the evolution of radiation-induced defects and the corresponding radiation hardening exhibit sooner, that appears to be because of the dose rate effect.

    更新日期:2018-11-14
  • Irradiation-induced amorphization in the zirconium suboxide on Zr-0.5Nb alloys
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-13
    J. Liu, G. He, J. Hu, Z. Shen, M. Kirk, M. Li, E. Ryan, P. Baldo, S. Lozano-Perez, C. Grovenor

    We report for the first time the observation of irradiation-induced amorphization of the zirconium suboxide formed during aqueous corrosion of Zr-0.5Nb alloys. High-resolution transmission electron microscopy results reveal amorphization of the hexagonal-ZrO suboxide under heavy ion irradiation at cryogenic temperatures. This irradiation-induced amorphization behaviour is discussed in relation to the arrangement of oxygen interstitials and the formation of stable superlattices. The sensitivity of the suboxide to irradiation damage can lead to phase changes and the accumulation of defects near the oxide/metal interface, which needs to be taken into account in the development of mechanistic models addressing radiation-assisted acceleration of corrosion rates in zirconium alloys.

    更新日期:2018-11-13
  • Corrosion assessment of 9Cr-1Mo steel in molten LiCl-KCl eutectic salt by electrochemical methods
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-13
    Ch. Jagadeeswara Rao, P. Venkatesh, S. Ningshen

    Ferritic chromoly steels are being considered for the process crucibles for the electrorefining of the pyrochemical reprocessing of spent metallic fuels in India. Pyrochemical reprocessing uses molten LiCl-KCl salt at 500–600 °C under an inert argon atmosphere for the electro-refining. The corrosion assessment of the 9Cr-1Mo steel in LiCl-KCl molten salt at 500 °C under inert argon atmosphere has been attempted using electrochemical measurements such as open circuit potential (OCP), linear polarization resistance and electrochemical impedance electrochemical impedance spectroscopy (EIS). The OCP of the sample was shifted towards noble direction during the total duration of 98 h. The linear polarization resistance, measured at regular intervals of exposure of 9Cr-1Mo steel to molten salt, increased with the increase of the exposure duration. The electrochemical impedance technique has been used to evaluate and monitor the molten salt-corrosion processes of the base metal. The EIS results showed the formation of a large semi-circle, with the characteristics of double loops capacitance attributed to the nature of intermittent oxide film over the surface. The surface oxide films are mostly composed of iron and chromium oxides confirmed by XRD and SEM-EDS analysis. It is also evidence from the EDS analysis that depletion and enrichment of Cr and Mo across the surface of the 9Cr-1Mo steel. This work indicated that the electrochemical techniques show considerable promise for the monitoring of high-temperature molten salt corrosion processes.

    更新日期:2018-11-13
  • On the role of intergranular carbides on improving the stress corrosion cracking resistance in a cold-worked alloy 600
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-13
    Zhao Shen, Junliang Liu, Koji Arioka, Sergio Lozano-Perez

    The mitigating effect introduced by intergranular Cr carbides on the stress corrosion cracking propagation of a cold-worked Alloy 600 has been firstly examined through high-resolution 3-dimensional (3D) sequential sectioning. High-resolution transmission electron microscope (TEM) and transmission Kikuchi diffraction (TKD) are used to reveal the underlying mechanisms contributing to the mitigating effect. Previously reported mechanisms contributing to the increased stress corrosion cracking resistance are evaluated and discussed. A new mechanism based on grain boundary migration inhibition and crack path deviation is proposed.

    更新日期:2018-11-13
  • The release behavior of boron and silicon from degraded absorber rods on core degradation during BWR severe accident
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-12
    A. Itoh, T. Sumita, M. Kajihara, Y. Kobayashi

    Activities of boron (B) and silicon (Si) in the Fe-Si-B liquid phase were evaluated by thermodynamic approach to clarify the effect of Si on the stable cesium (Cs) species formation at the temperature range from 1500 K to 2000 K. Although Si exists only as a minor alloying element of the stainless steel (SS) at the initial condition, Si may be controlling element when considering the chemical state of Cs in accidental conditions. The calculated activities were applied to the evaluation of chemical reactions of Cs-Si-O and Cs-B-O systems assumed in the Fukushima Daiichi Nuclear Power Station (FDNPS) unit 2. Consequently, production of 1.27 kg of the Cs-bearing particle at 2000 K was possibly found and it is considered to be specific phenomenon to the FDNPS unit 2 accident progression. Moreover, the following two findings were obtained that the increasing of Si resulted in an increase in activity of B in the melt, and Cs could not be released as CsBO2; instead, could be trapped on the surface of the SS-B4C condensed phase as the Cs-B-O species.

    更新日期:2018-11-12
  • Microstructural examination of zirconium alloys following in-pile creep testing in the HALDEN reactor
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-12
    Ken R. Anderson, Wade Karlsen, Mykola Ivanchenko, Jesse J. Carter, Richard W. Smith

    Post-irradiation examination (PIE) in the form of transmission electron microscopy (TEM) was used to characterize the microstructure of several specimens of Zircaloy-2 and Zircaloy-2 plus 1% Nb that had previously underwent in-pile creep testing in the HALDEN reactor. The purpose of the examination was to explore a microstructural basis for an apparent substantially increased rate of hardening over that observed with similar materials in the higher fast-flux environments of the ATR or HFIR, and to investigate the increased creep strength exhibited by the Nb-containing alloy relative to pure Zircaloy-2. The analysis of irradiation-induced defects indicated a higher than expected density, which was consistent with the observed high hardening rate. Modeling based on mean field rate theory suggests the lower neutron flux in the HALDEN reactor results in a higher fraction of irradiation-induced defects being available for sink (loop) nucleation and growth.

    更新日期:2018-11-12
  • Deuterium plasma driven permeation behavior in a Chinese reduced activation martensitic/ferritic steel CLF-1
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-12
    Hao-Dong Liu, Hai-Shan Zhou, Yu-Ping Xu, Yi-Ming Lyu, Lu Wang, Fang Ding, Guang-Nan Luo

    Deuterium (D) plasma-driven permeation (PDP) experiments for a Chinese reduced activation martensitic/ferritic steel CLF-1 have been performed. The effects of implantation flux, sample temperature and ion incident energy have been studied. The steady state PDP flux is found to be proportional to the square root of the implantation flux, indicating that the permeation takes place in recombination-limited regime for the upstream surface and diffusion-limited regime for the downstream surface (RD). The PDP flux and flux ratio (PDP flux/implantation flux) decrease with the increase of ion incident energy, which may be due to the enhancement of D recombination coefficient at the upstream surface.

    更新日期:2018-11-12
  • Assessing UO2 sample quality with μ-Raman spectroscopy
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-10
    K. Rickert, T.A. Prusnick, M.M. Kimani, E.A. Moore, C.A. Merriman, J.M. Mann

    μ-Raman spectroscopy of UO2 has demonstrated its value as a tool in nuclear forensics and for analyzing nuclear fuels, but its ability to evaluate UO2 sample quality has been relatively unexplored. To investigate this potential, a three-pronged study that uses Raman spectroscopy to analyze 1) three distinct qualities of UO2 single crystals with two different excitation lasers, 2) a high quality single crystal UO2 sample that is progressively damaged with a laser, and 3) a single crystal of UO2 grown on top of a ThO2 substrate was undertaken. The results demonstrate that the peak height ratio between the second order longitudinal optical phonon signal near 1145 cm−1 and the T2g peak located near 450 cm−1 is directly correlated with crystal quality when a 532 nm excitation source is used. Furthermore, this observed pattern relies upon the two most intense peaks that are present in the Raman spectra, making it more robust to lower signal-to-noise ratios than similar trends present in the peak height ratio between the first order longitudinal optical phonon signal near 565 cm−1 and the T2g peak. Both peak height ratios are more sensitive indicators of the crystal quality than the width ratios of the same signals or the associated wavenumber shift in the T2g peak location. The correlation between the 2LO and T2g peak heights and the overall crystal quality offers the potential to directly assess and compare the quality of different samples of single crystal UO2.

    更新日期:2018-11-10
  • Adsorption and diffusion mechanism of hydrogen atom on the Li2O (111) and (110) surfaces from first principles calculations
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-10
    Zhihong Yuan, Xianggang Kong, Shenggui Ma, Tao Gao, Chengjian Xiao, Xiaojun Chen, Tiecheng Lu

    The adsorption and diffusion of H atoms on (111) and (110) surfaces of Li2O lithium ceramics were investigated by the first principles method based on the density functional theory. The formation energies of these two surfaces were 0.546 J/m2 and 0.781 J/m2, respectively, which are close to those obtained by others. Combined with the adsorption energy, local electronic density of states, Bader charge analysis and charge density, we found that H atoms have three kinds of adsorption type on these two surfaces: physical adsorption, weak chemisorption and chemisorption. Comparing the diffusion barrier of H atoms on two surfaces, it is found that H atoms need to overcome over a barrier about 1.0 eV from top O atom to bridge site of two top Li atoms on (111) surface by CI-NEB. It is easier for the H atom to diffuse on the (110) surface with the diffusion barrier not to exceed 1.0 eV.

    更新日期:2018-11-10
  • Hydrogen blister formation in single crystal and polycrystalline tungsten irradiated by MeV protons
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-09
    I. Gavish Segev, E. Yahel, I. Silverman, A. Perry, L. Weismann, G. Makov

    Single and polycrystalline tungsten samples were irradiated with 2.2 MeV protons at Soreq Applied Research Accelerator Facility (SARAF). Hydrogen blisters were obtained for both single crystal and polycrystalline samples, elucidating the role of grain boundaries in blister formation. The effect of temperature and flux on the critical formation dose for blisters and on their dimensions was studied. It was found that for single crystals, the critical formation dose is one order of magnitude higher than for polycrystalline tungsten at high temperature irradiation conditions. Upon reducing the irradiation temperature to ambient, the critical dose for formation of blisters in single crystals was reduced by a factor of three while in polycrystalline tungsten there was no significant change with temperature, thus indicating the role of grain boundaries in blister formation. Larger blisters were obtained in single crystals than in polycrystalline tungsten at ambient temperature conditions, identifying the grain boundaries as a preferential additional hydrogen trap. The height to area ratio of the blisters is found to be strongly temperature dependent and only weakly dependent on irradiation flux for both single and polycrystalline samples.

    更新日期:2018-11-09
  • Infrared spectroscopy of ion tracks in amorphous SiO2 and comparison to gamma irradiation induced changes
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-09
    M. Karlušić, M. Škrabić, M. Majer, M. Buljan, V. Skuratov, H.K. Jung, O.Gamulin, M. Jakšić

    Ion track formation in amorphous SiO2was investigated using infrared spectroscopy. For comparison, one set of samples was also irradiated using 1.25 MeV gamma rays. An increase of 1044 cm−1 peak and decrease of 1078 cm−1 peak was observed in all cases. Experimental results were analysed using an analytical thermal spike model and non-standard model parameters were found. This finding is attributed to the amorphous structure of the material.

    更新日期:2018-11-09
  • Short communication: The effect of cooling rate and grain size on hydride microstructure in Zircaloy-4
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-08
    R. Birch, S. Wang, V. Tong, B. Britton

    We explore the distribution, morphology and structure of zirconium hydrides formed using different cooling rates through the solid state Zr+[H] → Zr + hydride transus, in fine and blocky alpha Zircaloy-4. We observe that cooling rate and grain size control the phase and distribution of hydrides. The blocky alpha (coarse grain, > 200 μm) Zircaloy-4, has a smaller grain boundary area to grain volume ratio and this significantly affects nucleation and growth of hydrides as compared to fine grain size (∼11 μm) material.

    更新日期:2018-11-08
  • Effect of high temperature annealing on the microstructure and thermal shock resistance of tungsten coatings grown by chemical vapor deposition
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-07
    Youyun Lian, Fan Feng, Jianbao Wang, Xiang Liu, Jiupeng Song, Yingmin Wang, Zhe Chen, Jiming Chen

    Thick tungsten (W) coatings were prepared by chemical vapor deposition (CVD) at a rapid growth rate of 0.6 mm/h. Annealing experiments have been performed in a relatively wide temperature range from 1200 °C to 2300 °C for 3 h to examine the effect of high temperature exposure on the microstructure, micro-hardness and thermal shock resistance of the CVD-W coatings. The columnar grain structures with a preferred <001> orientation of the CVD-W coatings are well preserved up to 2300 °C. Bubbles of polyhedral forms formed at the grain boundaries of the sample annealed at 2300 °C. The dislocation structures were characterized by a high density of dislocation tangles and a low density of long and straight screw dislocations in the samples annealed below and above 1900 °C, respectively. The micro-hardness of the CVD-W coatings decreased with the increase of annealing temperature. The polished CVD-W coating surfaces exhibited high thermal shock resistance to ELM-like thermal loadings. Surface crack formation was suppressed at elevated temperature except for the sample annealed at 2300 °C.

    更新日期:2018-11-07
  • Energetics and kinetics of metal impurities in the low-temperature ordered phase of V2C from first-principles calculations
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-05
    B.J. Demaske, A. Chernatynskiy, S.R. Phillpot

    The site preference of Fe, Ni, Ce, Nd and U impurities and their migration behaviors in the low-temperature ordered orthorhombic phase of V2C are calculated using density functional theory. It is found that all impurities prefer the substitutional vanadium site over the octahedral interstitial site. The energy required to incorporate an impurity into the lattice increases with increasing impurity size. Binding between an impurity at the substitutional vanadium site and a neighboring vanadium vacancy is shown to be much stronger for large impurities, U, Ce and Nd, than for small impurities, Fe and Ni. The strong binding can be attributed to the openness of the V2C structure, which allows large impurity atoms to relax by shifting towards neighboring empty octahedral sites. The proximity of each impurity-vacancy pair to nearby carbon interstitial atoms leads to a high degree of anisotropy in binding strength with binding energies differing by > 2 eV for the same impurity type. Small impurities, Fe and Ni, have negative impurity volumes at the substitutional vanadium site, so binding with neighboring vanadium vacancies is much weaker. Migration barriers for direct exchange of an impurity with a vanadium vacancy are calculated for Fe and Ni. On average migration barriers for Ni are smaller than those for Fe.

    更新日期:2018-11-05
  • First-principles study of surface properties of uranium silicides
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-05
    Zhi-Gang Mei, Yinbin Miao, Linyun Liang, Abdellatif M. Yacout

    Uranium silicides are currently under investigation as accident tolerant fuels for light water reactors because of its high uranium density and high thermal conductivity. Surface energy as an important material property is required for modeling of gas bubble behavior in nuclear fuels using mesoscale approaches, such as phase field and rate theory methods. However, there is no such information available for uranium silicides from either experiment or theory. To this end, we study the surface properties of two uranium silicide compounds U3Si2 and U3Si using first-principles calculations. Of the low-index facets of tetragonal U3Si2 and U3Si, we study a total of 13 surfaces up to a maximum Miller index of 3. From the calculated surface energies, the equilibrium single crystal shapes of U3Si2 and U3Si are obtained using Wulff construction. The dominant surface orientation, surface area weighted surface energy and surface anisotropy are predicted. The obtained surface properties of U3Si2 and U3Si are crucial for an accurate description of the morphology of fission gas bubbles in uranium silicide fuels.

    更新日期:2018-11-05
  • Recrystallization kinetics of cold-rolled U-10 wt% Mo
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-02
    William E. Frazier, Shenyang Hu, Nicole Overman, Ramprashad Prabhakaran, Curt Lavender, Vineet V. Joshi

    Samples of U-10 wt%Mo (UMo) alloy were thermomechanically processed and annealed at 600 °C and 700 °C for periods lasting up to 8 h. Annealed microstructures were examined using electron backscattered diffraction in order to estimate the completeness of recrystallization. Hardness measurements of the UMo samples were taken to further qualify the progress of recovery and recrystallization. This allowed us to evaluate the recrystallization activation energy and the recrystallization kinetics with a Johnson-Mehl-Avrami-Kolmogorov equation. The results show that the activation energy of recrystallization is approximately 100.6 ± 23.1 kJ/mol. The mechanisms active in the recrystallization of cold-rolled U-10 wt% Mo are discussed in the context of these results.

    更新日期:2018-11-05
  • The impact of azimuthally asymmetric carbon deposition upon pellet-clad mechanical interaction in advanced gas reactor fuel
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-02
    T.A. Haynes, V. Podgurschi, M.R. Wenman

    Cracked nuclear fuel pellets were modelled in the r-θ plane with an azimuthally varying clad surface temperature boundary condition using the PELICAN set of fuel performance models for the commercial finite element software, Abaqus. The temperature boundary condition was assumed to represent heat transfer impairment due to an azimuthally asymmetric carbon deposit on advanced gas-cooled reactor pins. The model predicts the radial and azimuthal displacement of the idealised fuel fragments, together with the resulting elastic, creep and plastic strains in the cladding. These were compared to simulations assuming a uniformly hot or cold boundary condition. Apart from a short period during the return to power from reduced power (70%) operation and outages, the hoop stress in the simulation with an azimuthally varying clad surface temperature was bounded by that of models with a uniform hot or cold surface temperature. The reduced stress was proposed to be due to the greater ability of the fuel fragments to relocate in order to accommodate changes to the power level. As a result, the creep strains in the model with an azimuthally varying clad surface temperature were lower than assuming either a uniform hot or cold boundary condition.

    更新日期:2018-11-05
  • First-principles study of thermophysical properties of interaction layer products in U-Mo/Al dispersion fuel
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-02
    Zhi-Gang Mei, Abdellatif M. Yacout

    Thermophysical properties of interaction layer (IL) formed in U-Mo/Al dispersion fuel are important for the evaluation of its impact on the fuel performance. Binary and ternary phases observed in the IL, however, are not well studied yet. Recently we predicted the thermophysical properties of the binary U-Al compounds in U-Al system using density functional theory (DFT) calculations. In this paper, we investigated the structural, elastic, electronic, vibrational and thermodynamic properties of the ternary U-Mo-Al alloy compounds formed in IL, including (U0.75,Mo0.25)Al3, U6Mo4Al43 and UMo2Al20, using DFT calculations. The polycrystalline aggregate properties of these compounds were obtained from the predicted single crystal elastic constants. The calculated electronic density of states confirm that all of the compounds exhibit typical metallic behavior with the majority states at the Fermi level dominated by U 5f electrons. Using quasi-harmonic approximation, we predicted the thermodynamic properties of these compounds by including both electronic and lattice contributions. We also investigated the effect of density on the thermodynamic properties of (U0.75,Mo0.25)Al3 phase. The current results are expected to be helpful to the modeling of the fuel performance of U-Mo/Al dispersion fuel and the thermodynamic modeling of the ternary U-Mo-Al system.

    更新日期:2018-11-05
  • Assessment of Te as a UZr fuel additive to mitigate fuel-cladding chemical interactions
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-02
    Yi Xie, Jinsuo Zhang, Michael T. Benson, James A. King, Robert D. Mariani

    Tellurium was investigated as a potential additive to metallic fuel (e.g. U10Zr) to mitigate fuel-cladding chemical interaction (FCCI). A primary cause of FCCI is the lanthanide fission products migrating to the fuel periphery and interacting with cladding, Te therefore was used to bind the lanthanides into stable compounds in the U10Zr. The present work investigates the microstructures of as-cast and annealed U10Zr4Te and U10Zr4Te4Ce alloys (weight percent) with scanning electron microscope and transmission electron microscopy. Tellurium was found to bind Ce and form CeTe compound, which will mitigate lanthanides migration to the cladding.

    更新日期:2018-11-05
  • Thermal radiative properties of electron-beam-melted and mechanically alloyed V-4Cr-4Ti based alloys between 200 and 750 °C
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-02
    T. Echániz, I. González de Arrieta, R. Fuente, I. Urcelay-Olabarria, J.M. Igartua, N. de la Pinta, W. Ran, H. Fu, J. Chen, P.F. Zheng, M.J. Tello, G.A. López

    The directional spectral emissivities of two V-4Cr-4Ti family alloys, candidate structural materials for fusion first wall/blanket applications, were measured between 200 °C and their working temperatures (700–750 °C), with and without a high-temperature treatment. Besides showing the typical metallic behavior, an increase in the emissivity after the heat treatment (1000–1200 °C) was observed in both alloys. This has been attributed to several microstructural changes, which show the important role of microstructure in the thermal radiative properties of these alloys. In order to explain these mechanisms, the samples were analyzed using electron microscopy and X-ray diffraction. These measurements revealed differences in grain size, composition of the main phase and amount and distribution of dispersed secondary phases. X-ray diffraction and X-ray photoelectron spectroscopy were also used in order to check the extent of oxygen penetration. The results of directional spectral emissivity measurements were integrated to calculate the total hemispherical emissivity, the key heat transfer parameter in the high-temperature high-vacuum environments of fusion reactors. It is observed that the strategy of mechanical alloying with oxide and carbide dispersion to improve the mechanical properties also translates into an enhancement of the radiative refrigerating capability of these alloys.

    更新日期:2018-11-05
  • Synthesis and characterization of phosphate-based glass-ceramics for nuclear waste immobilization: Structure, thermal behavior, and chemical stability
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-03
    Jinfeng Liu, Fu Wang, Qilong Liao, Hanzhen Zhu, Dongsheng Liu, Yongchang Zhu

    Phosphate-based glass-ceramics were prepared by a traditional glass melt-quenching method in a high-temperature furnace. The effects of the ZrO2 substituted for Na2O on the crystal phase, thermal behavior, structure, and chemical stability of the prepared glass-ceramics were investigated in detail. X-ray diffraction (XRD) analysis showed that NaZr2(PO4)3 is the main crystalline phase of all the studied samples. ZrP2O7 and FePO4 respectively appeared when the ZrO2 substituted for Na2O reached or exceeded 8 mol% and 12 mol%. FTIR and Raman spectra showed that the network structure of the studied glass-ceramics consisted mainly of orthophosphate units, pyrophosphate units, and a small amount of metaphosphate units and [BO4] units. For the samples containing 10 mol% ZrO2 or less, the normalized leaching rates of Zr (LRZr), Fe (LRFe), Na (LRNa), and P (LRP) remained low (2.5 × 10−6 g m−2 d−1, 3.3 × 10−6 g m−2 d−1, 2.5 × 10−3 g m−2 d−1, and 6.2 × 10−4 g m−2 d−1, respectively) after immersion in deionized water at 90 °C for 28 days. The obtained results suggest that the traditional glass making process may potentially be applicable to the synthesis of some phosphate-based glass-ceramics for immobilizing nuclear waste.

    更新日期:2018-11-05
  • Comparison of irradiation tolerance of two MAX phases-Ti4AlN3 and Ti2AlN
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-03
    Tengfei Yang, Chenxu Wang, Wulong Liu, Shaoshuai Liu, Jingren Xiao, Qing Huang, Yugang Wang, Steven J. Zinkle

    The microstructures of two MAX phases, Ti2AlN and Ti4AlN3, irradiated at room temperature with 70 keV He ions over a wide fluence range have been studied to reveal the dependence of MAX phase irradiation tolerance on structure. Grazing incident x-ray diffraction (GIXRD) reveals that a new set of diffraction peaks appears after ion irradiation for both MAX phases and the positions of diffraction peaks gradually shift towards lower 2θ values with increasing fluence. Transmission electron microscope characterizations show both MAX phases experience a similar structure transformation at low fluences, forming an intermediate structure. The d-spacings derived from electron diffraction analysis of intermediate structures are basically consistent with that of new peaks in GIXRD. At high fluences a nano-twinned microstructure is found in Ti2AlN, but no such microstructure is observed in Ti4AlN3. Based on the simulation results of electron diffractions and the irradiation responses of other MAX phases, the structure characteristics and formation mechanisms of the intermediate structure are derived and the discrepancy in the phase stability of both MAX phases at high fluences is discussed.

    更新日期:2018-11-05
  • 更新日期:2018-11-02
  • A universal COMB potential for the whole composition range of the uraniumoxygen system
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-01
    Yangzhong Li

    An empirical molecular dynamics potential in the Charge-Optimized Many-Body (COMB) formalism that covers the whole uranium-oxygen composition range has been developed. Extended from a previous potential for uranium metal, this universal UO potential is able to model more than 20 phases of uranium oxides. The potential's flexibility, accuracy and transferability have been fully verified by rigorous testing and comparison with ab-initio calculations and experimental measurements. It is shown to be one of the most versatile and high-quality UO2 potentials, the first potential for U4O9, U3O7 and UO3, and the first usable U3O8 potential. Many important properties of major oxides in the UO phase diagram have been calculated and critically reviewed, including the cohesive energy, formation/reaction energies, lattice parameters, elastic constants, bulk/shear moduli, and energies for non-stoichiometric point defects and stoichiometric defects pairs. Due to its special design and parameterization process, this UO potential is shown to outperform all other existing ones by either obtaining higher accuracy for many of these quantities, or exclusively being able to calculate some of them. The successfully development of this potential provides a useful, reliable and convenient tool for molecular dynamics simulations that are previously impossible or unreliable to do for many materials in the UO system. Correction for several published oxide structures is also included in the appendix.

    更新日期:2018-11-02
  • Deformation analysis of SiC-SiC channel box for BWR applications
    J. Nucl. Mater. (IF 2.447) Pub Date : 2018-11-01
    G. Singh, J. Gorton, D. Schappel, N.R. Brown, Y. Katoh, B.D. Wirth, K.A. Terrani

    Silicon carbide fiber-reinforced silicon carbide matrix (SiC-SiC) composites are being considered as components in light water reactor cores to improve accident tolerance, including channel boxes and fuel cladding. In the nuclear reactor environment, core components like a channel box will be exposed to neutron and other radiation damage and temperature gradients. To ensure reliable and safe operation of a SiC-SiC channel box, it is important to assess its deformation behavior under in-reactor conditions including the expected neutron flux and temperature distributions. In particular, this work has evaluated the effect of non-uniform dimensional changes caused by spatially varying neutron flux and temperatures on the deformation behavior of the channel box over the course of one year. These analyses have been performed using the fuel performance modeling code BISON and the commercial finite element analysis code Abaqus, based on fast flux and temperature boundary conditions that have been calculated using the neutronics and thermal-hydraulics codes Serpent and CTF, respectively. The dependence of dimensions and thermophysical properties on fast flux and temperature has been incorporated into the material models. These initial results indicate significant bowing of the channel box with a lateral displacement greater than 6.5 mm. The channel box bowing behavior is time dependent and driven by the temperature dependence of the SiC irradiation-induced swelling and the neutron flux/fluence gradients. The bowing behavior gradually recovers during the course of the operating cycle as the swelling of the SiC-SiC material saturates. However, the bending relaxation due to temperature gradients does not fully recover and residual bending remains after the swelling saturates in the entire channel box.

    更新日期:2018-11-02
Some contents have been Reproduced with permission of the American Chemical Society.
Some contents have been Reproduced by permission of The Royal Society of Chemistry.
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