查看更多>>摘要：A 30 MeV Ni-58(5+) ion beam was used to irradiate fine-grained graphite grade IG110 and ultrafine-grained graphite grade G1 at 400 degrees C to study their microstructure evolution under irradiation. Taking advantage of the depth dependence of the damage rate and cumulative damage of ion irradiation, the microstructure change exploration of graphite with multiple damage levels within a single fluence specimen was achieved by characterizing the sample cross-section with HR-TEM and micro-Raman. The Raman 2D maps of the cross-sections of the graphite samples irradiated with various fluences and displacement damages up to similar to 18 displacements per atom were analyzed. Evidenced by the saturation of the intensity ratio of graphite Raman D and G band (I-D/I-G), the irradiation damage and annealing equilibrium was observed. Moreover, I-D/I-G in combined with the full width at half maximum of the G band (FWHM (G)), shows an inverse evolution compared with the graphitization process, which is also supported by the HR-TEM observation. Demonstrated by the increasing rate of I-D/I-G with FWHM (G) and the saturation I-D/I-G, the microstructure changes of fine-grained graphite IG110 and ultrafine-grained graphite G1 was distinguished. (c) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：S/TEM imaging has shown that the hydrides precipitated in Zr-2.5Nb pressure tube alloy under externally applied hoop tensile stress or under stress-free conditions show different morphologies. When precipitating under stress, at the bulk scale the hydrides tend to align towards the radial direction, forming so-called 're-oriented' or radial hydrides. At the grain level, the observed re-oriented hydrides consisted of hydride platelets growing along the tube radial direction. In contrast, the hydrides precipitated under stress-free conditions were orientated along the transverse direction, and consisted of platelets that had grown along the tube transverse direction, and exhibited almost full transformation of the parent Zr grain. Upon heating, TEM samples where both transverse and radial oriented platelets were present, the radial platelets were shown to be more resistant to heating, i.e., dissolving at a higher temperature than the transverse hydrides. The migration of the hydride/Zr phase boundary was observed to follow different directions during dissolution, depending on the morphology of the individual hydrides. Lastly, the texture of the Zr grains where hydrides precipitated was analyzed by electron diffraction. Results showed that hydrides preferred to form in grains with (0 0 02) pole aligned in/or close to the TD (transverse direction)-RD (radial direction) plane when external hoop tensile stress was applied, and the applied stress showed a profound impact on where hydrides precipitated within the available population of differently oriented Zr grains. (c) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：The loop evolution in Mo-5 wt.%Re alloy was studied by in-situ TEM observation under 400 keV Fe + irradiation at 700 degrees C and 800 degrees C. Loop initiation always existed during the whole irradiation process. The gradual disappearance of dislocation line to grain boundary and the change of loop habit plane from [101] to [211] were in-situ observed, and the changed loop acting as an intermediate bridge led to the merging of three loops. Three main ways causing loop disappearance including the aggregation between loops, the absorption by strong defect sinks and the influence of surrounding loops were obtained from in-situ experiment. With the increase of irradiation dose, many loops formed in coarse grains but few in nanocrystals, and the areal density of dislocation lines almost kept stable in the grain with many preexisting dislocation lines. The variation trends and relationships of the average size and volume number density of loops with irradiation temperature, irradiation dose and sample thickness were obtained. At the same irradiation temperature, the thicker the thickness of TEM foil was, the higher the yield strength increasement would be. Meanwhile, for the same sample thickness, higher temperature would reduce the irradiation hardening effect. The in-situ observation of microstructure evolution will be helpful to deeply understand the irradiation damage behavior of Mo-Re alloy. (c) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：Damage behaviors of ITER grade tungsten (W) induced by repetitive thermal loads have been investi-gated using an electron beam device EBMP-30. The metallographic observation, backscattered electron scanning electron microscopy, electron backscattered diffraction, atomic force microscopy analysis and tensile tests were carried out after the repetitive heat loads. Results indicate that there are negligible ef-fects on the microstructures, surface morphologies and mechanical properties with the absorbed power densities (APD) lower than 10 MW/m(2). While serious damages including the severe surface roughness with protruding structures along grain boundaries, and significant grains coarsening with the evolution of the grain shape occur undesirably due to the full recrystallization during the repetitive 30 MW/m(2) heat loads. Furthermore, after exposed to an APD of 30 MW/m(2 )with 50 heat loads, the tensile proper-ties deteriorate dramatically, resulting in a limited ultimate tensile strength of only 328 MPa and even a zero total elongation at 300?. The relationships between the thermal loads, microstructures evolution and degradation of mechanical properties have been analyzed.(c) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：Molten Salt Reactors (MSRs) are envisioned as a potential pathway to safer, more economical nuclear electricity generation and supply of industrial heat. MSRs under consideration today are either solidfueled salt-cooled designs or liquid-salt-fueled designs with chloride or fluoride based salts. A significant knowledge gap exists in the data for the fundamental properties relevant to fuels and coolants for MSRs that needs to be addressed in order to expedite the technical readiness level of the MSR design concepts. With the rapid development and improvement of computational materials science, computational methods such as Density Functional Theory (DFT) calculations and ab initio Molecular Dynamics (AIMD) simulations are widely used as an effective and reliable tool to investigate the atomic interaction in materials. In this article, the density of the LiCl-KCl system was determined via AIMD calculations and verified using new experimental analyses. AIMD was further utilized to calculate the compressibility, heat capacity, enthalpy of mixing, and Gibbs free energy of mixing. This work spans a wider range of compositions and temperatures than have previously been explored computationally for this pseudo-binary system and provides the basis for further advanced thermophysical property evaluation utilizing AIMD methods. Published by Elsevier B.V.
Spano, Tyler L.Hunt, RodneyKapsimalis, Roger J.Niedziela, Jennifer L....
9页
查看更多>>摘要：epsilon-UO3 is an exotic polymorph in the uranium trioxide system with an undetermined crystal structure and limited optical vibrational spectroscopic data. To improve understanding of this compound, we synthesize and investigate the crystal structure and optical vibrational spectra of epsilon-UO3. Infrared spectra collected for epsilon-UO3 are in good agreement with previously published results, and our studies extend the available data into the low-energy (600-100 cm(-1)) regime. For the first time, Raman spectra are presented for epsilon-UO3 using both 785 and 532 nm excitation wavelengths. Previous reports suggest an impurity phase may be present in epsilon-UO3 produced by calcination of U3O8; however, spectral center-of-mass calculations, principal component analyses, and Raman spectroscopic mapping employed to investigate this possibility indicate that the product of U3O8 calcined in O-3(g) in this work is likely phase-pure. A possible novel structure solution for epsilon-UO3 is determined via Rietveld refinement of powder X-ray diffraction data and is triclinic, P-1, with a = 4.01 angstrom, b = 3.85 angstrom, c = 4.18 angstrom, and alpha = 98.26 degrees, beta = 90.41 degrees, gamma = 120.46 degrees (R-wp = 8.30%). The asymmetric unit of epsilon-UO3 consists of U(VI) in hexagonal bipyramidal coordination with displaced equatorial oxygen. Further analysis reveals that the structure of epsilon-UO3 is best described by a 2 x 1 x 2 supercell structure in P-1 with a = 8.03 angstrom, b = 3.86 angstrom, c = 8.37 angstrom with alpha = 98.26 degrees, beta = 90.41 degrees, and gamma = 120.46 degrees, although a higher-symmetry structure is possible. Optical vibrational spectroscopic and structural measurements of epsilon-UO3 presented here furthers our understanding of this complex uranium oxide and clarifies the origin of reported structural similarity to U3O8. (C) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：As one of the most promising first wall/blanket structure materials in fusion reactors, oxide dispersion strengthened (ODS) ferritic steel has been extensively studied in past decades. The grain size of ODS steels is often between 200and 1000 nm, called ultrafine-grained (UFG). Refining their grain size, if possible, should further enhance their radiation tolerance. In the present work, we report on a novel zirconium-doped nanocrystalline (NC) 14YWTZ ODS steel composed of a ferritic matrix with an average grain size of 50 nm and high-density oxide nanoprecipitates with an average diameter of 3.3 nm. Both NC and UFG 14YWT ODS steels were irradiated with helium ions at 450 degrees C. Abnormal lattice shrinking and narrowing of X-ray diffraction peaks are found in irradiated NC ODS steel. The NC ODS steel has an extremely high sink strength of -3 x 10 16 m(-2) , which is mainly contributed by grain boundaries and effectively inhibits the aggregation of He atoms and the growth of He bubbles. The bubble size, void swelling, and irradiation hardening in NC ODS steel irradiated at a high dose, when compared to those in UFG ODS steel, are significantly smaller. The underlying mechanisms for the high irradiation tolerance in the NC ODS steel are discussed. This work provides an approach to further enhancing the radiation resistance of conventional UFG ODS steels by refining their grain size to nanoscale dimensions. (c) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：Two tristructural isotropic (TRISO)-coated nuclear fuel particles were examined by electron probe micro-analysis (EPMA) as part of the Advanced Gas Reactor program. The compacts' average irradiation temper-atures ranged from approximately 1260 to 1290 degrees C. One particle was examined in the as-irradiated con-dition, while the other was subject to 1600 degrees C post-irradiation safety testing. This study was undertaken to test a newly-developed EPMA technique to determine fission product masses in TRISO particles on a layer-by-layer basis, and to compare fission product distributions between an as-irradiated and safety -tested particle. Fission product concentration profiles were collected along two radii in each particle, with measured concentrations used to compute the fission product mass in each TRISO particle layer. These measured masses were then compared to those predicted from ORIGEN modeling calculations. Data col-lected from these measurements show that for these two particles, masses determined via EPMA were within +/- 20% of the calculated masses for the rare-earth elements, Mo, Zr, Cs, I, and Pd. Elements that tend to be less homogeneously distributed include Sr, Te, Eu, Ag, and possibly Ba. Measured Ag masses differed by more than 40% from the calculated mass. Lanthanides other than Eu remain primarily within the fuel kernel in the as-irradiated particle but in the safety-tested particle these element masses were divided approximately equally between the kernel and kernel periphery. In both particles, the majority of Sr and Eu accumulated in the carbon-rich kernel periphery, although in the safety-tested particle, Sr and Eu accumulated farther from the fuel kernel than occurred with irradiation alone. A greater mass fraction of mobile elements, such as Cs and I accumulated in the buffer and IPyC in the safety-tested particle as compared to the as-irradiated particle. When fully developed and tested, this mass balance approach to TRISO particle analysis has the potential to provide insight into fuel behavior.(c) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：Cu piping previously used to carry process air was harvested from the now shutdown National Research Universal reactor at the Canadian Nuclear Laboratories site in Chalk River, Ontario. Metallurgical and mi-croscopy examinations found no quantifiable differences between irradiated and non-irradiated sections of the pipe. This provides evidence for the safety case for the use of the Cu-coated used fuel container proposed by the Nuclear Waste Management Organization for the permanent disposal of high-level nu-clear waste. (c) 2021 Elsevier B.V. All rights reserved.
查看更多>>摘要：The complete and safe disposal of large irradiated graphite waste from nuclear reactors has attracted worldwide attention. The core target for the treatment of irradiated graphite is to decontaminate the trace radioactive elements. The reviewed investigation mainly focused on the performance of the decon-tamination method of irradiated graphite based on the current available literature. Survey to fabrication, structure and radioactivity of nuclear graphite was first introduced. Several common treatment technolo-gies (incineration, thermal treatment, electrochemical or chemical treatment, plasma technique, molten salt flameless oxidation and self-sustaining high-temperature synthesis) were summarized. The recom-mendations, standpoints and potential research on the decontamination method of irradiated graphite were highlighted. The work presented would contribute to the information and experience to build de-contamination methods and develop new technologies.(c) 2021 Elsevier B.V. All rights reserved.
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