查看更多>>摘要：A generic rig with a large range of possible irradiation conditions has been successfully developed for the BR2 material test reactor at SCK CEN. The rig contains five capsules in which active temperature control up to similar to 300 degrees C is possible, and irradiation temperature and neutron dose can be chosen independently in each capsule. Each capsule can be lifted out of the core region of BR2 even during the operation of BR2 when the dose for the capsule reached the requested value. This results in material irradiation experimental data for which the irradiation dose is the only single parameter. (C) 2022 The Author(s). Published by Elsevier B.V.
Lin, YushaMcFadzean, CharlesHumphry-Baker, Samuel A.
10页
查看更多>>摘要：Tungsten borides are candidate radiation shielding materials for compact fusion reactors. However little is known about their performance when oxidised at high temperatures in the case of an accident combining loss of coolant and vacuum. Candidate materials containing 0, 16, 30, and 50 at.% B were exposed to air at 60 0-1100 degrees C in a thermogravimeter and the kinetic rate constants were evaluated. The corresponding oxide scales were characterised by X-ray diffraction and electron microscopy. The boron containing materials showed parabolic kinetics, more protective behaviour than tungsten, and an improvement in oxidation resistance with increasing boron content. For example, the rate constant was a factor of 600 lower for 50 at.% B vs. pure W at 1000 degrees C. In 16 and 30 at.% B materials the scale formed a microstructure of interpenetrating B2O3 and WO3. In the 50 at.% B material, two distinct layers formed, with a B2O3-rich layer on the surface, and fine WO3-based layer beneath. The protective scale was disrupted in all materials at 1000-1100 degrees C, depending on boron content. The disruption resulted in a transition to linear kinetics and can be explained by the B2O3 evaporation rate exceeding its formation rate. Encouragingly, all tungsten borides have superior oxidation resistance compared to pure tungsten, suggesting favourable accident tolerance. (C) 2022 The Author(s). Published by Elsevier B.V.
查看更多>>摘要：Silicon carbide (SiC) cladding is being studied for its potential application in light water reactors for improved accident tolerance. SiC-SiC composite material undergoes irradiation-induced, temperature dependent swelling. The circumferentially non-uniform neutron flux and power generation in the fuel rod, will produce circumferentially non-uniform swelling and thermal expansion, ultimately leading to bending of the cladding. This bending will cause interaction between the cladding and the spacer grids and can be detrimental to the structural integrity of both the cladding and spacer grids. In this study we have analyzed the bending behavior of the cladding, the stress development in the cladding and the effects of spacer grids. The sensitivity of the cladding stress to the spacer grid material, extent of circumferential variation in power have been also evaluated. The results indicate that the spacer grids do not significantly affect the cladding stresses. The stress generated in the spacer grids due to its interaction with the cladding is found to be well below the strength of the spacer grid material. The sensitivity studies showed that changing the spacer grid material have negligible impact on the stresses in both cladding and spacer grid, while the extent of circumferential variation in power can lead to significant increase in the stresses in spacer grid. (c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：The influence of the transmutation element rhenium (Re) content with 3 and 25 wt.% on deuterium (D) permeation in tungsten (W) was investigated. It was found that the addition of Re increased D diffusion coefficient and decreased D permeability, and the D diffusion coefficient of W-3%Re was approximate to that of W-25%Re while the D permeability declined with the increase of Re content. The D retention behavior in the W-Re alloys have been studied after D plasmas exposure employing the Material and Plasma Evaluation System (MAPES) in Experimental Advanced Superconducting Tokamak (EAST). As for the higher D solution energy in W-Re alloys, the total D inventory in W was significantly decreased by Re addition. (c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：Supercritical water-cooled reactor candidate materials are typically corrosion-resistant alloys whose mass changes in supercritical water fluctuate around zero. A previous interlaboratory corrosion experiment exercise revealed large scatter in mass change results between laboratories. Here, we reduced systemic differences between laboratories by unilateral preparation of coupons and unilateral chemical cleaning. Type 310S stainless steel and Alloy 800HT test coupons were exposed for 10 0 0 h to 550 degrees C, 25 MPa, deaerated water. Average mass loss for 310S and 800HT was 38 +/- 26 and 51 +/- 31 mg/dm(2), respectively. Differences in mass transfer, galvanic and local corrosion, and average coupon temperature may explain the poor reproducibility. (C)& nbsp;2022 Published by Elsevier B.V.
查看更多>>摘要：Chemical disordering is one feature of damage that occurs in irradiation environments and is common in ceramic compounds. This paper reviews irradiation-induced chemical disordering in SiC-based materials at elevated temperatures. The results were obtained using advanced analytical tools to characterize atomistic defects and to evaluate the chemistry of defect clusters and homonuclear bond structures. This paper demonstrates that chemical disordering is crucial to understanding microstructural stability and establishing a mechanistic modeling of microstructural evolution of irradiated SiC.(c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：Reactor pressure vessels (RPVs) are one of the main barriers against nuclear accidents, therefore good toughness is indispensable for keeping the integrity of RPVs to prevent the leakage of radioactive substances. It is well known that high temperature and large thermal gradient of nuclear accidents would lead to embrittlement or even rupture of RPVs. However, how damages originate in severe nuclear accidents is still indistinct. This study is aimed to reveal the embrittlement mechanism under a large temperature gradient in quenched RPV steels. It is found that certain strengthening as well as remarkable embrittlement occurs after quenched, further investigations find it probably derive from the generation of plenty of Moire fringes and dislocation networks, which are supposed to be generated by shearing of crystal planes. This paper reports the deterioration phenomenon caused by a large thermal gradient, reveals the origins remarkable embrittlement and indicates that extra attention must be aroused to take this thermal-stress-induced embrittlement into consideration in nuclear accidents analysis.(c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：From a safety standpoint, it is important to understand the behavior of tritium in the coolant of DT fusion reactors. The high-temperature and high-pressure water is expected to be used as a coolant in JA DEMO. Therefore, it is necessary to deepen the understanding of the tritium transfer from the primary cooling water to the secondary cooling water in a heat exchanger. In this study, a double-tube permeation device consisting of Inconel 600 tube, which is a heat exchanger material, and SS316 tube, was assembled and the tritium permeation from tritiated water to water at 300 degrees C and 17 MPa was observed. Remarkable tritium permeation was observed after around 17 days as the cumulative heating time. Gaseous tritium (HT and T 2 ) was detected in the gas phase of the tritiated water side after the permeation experiments. This suggests that a part of tritium generated in the metal oxidation reaction dissolved in the Inconel and diffused and permeated. With assuming that the diffusion process of tritium through the Inconel was the rate controlling step of permeation, the permeability of tritium through the Inconel tube from the tritiated water derived from the experimental data was 3 orders of magnitude smaller than that obtained from the hydrogen isotope permeation experiments from gas phase to gas phase. (c) 2022 Published by Elsevier B.V.
Herrmann, Steven D.Zhao, HaiyanBawane, Kaustubh K.He, Lingfeng...
12页
查看更多>>摘要：Given a growing interest in uranium salts for pyrochemical processing of used fuel and uranium-fueled molten salt reactors, the synthesis of uranium trichloride in alkali-metal chloride media was investigated in a series of four experiments. Specifically, uranium metal powder and uranium hydride powder were prepared and separately blended with ammonium chloride and lithium chloride - potassium chloride eutectic in two runs, while the same powders were separately blended with ammonium chloride and sodium chloride in two additional runs. Each of the lithium chloride - potassium chloride containing blends was slowly heated to 923 K, while those containing sodium chloride were heated to 1123 K. During each heat up, the ammonium chloride sublimed into gaseous ammonia and hydrogen chloride, leading to the chlorination of uranium metal or uranium hydride and the formation of molten salt solutions of the respective chlorides. Experimental conditions were incorporated in the runs to promote formation of uranium trichloride over uranium tetrachloride in the respective media. Molten samples of each run product were taken and characterized via chemical analyses, diffractometry, and microscopy. The final products from each run were dark dense ingots of the respective salt systems with uranium concentrations ranging from 44 to 51 wt%. Chemical analyses and diffractometry identified the predominant presence of uranium trichloride in these systems; however, a possible minor presence of uranium tetrachloride could not be conclusively dismissed.(c) 2022 Elsevier B.V. All rights reserved.
查看更多>>摘要：Two CrFeMnNi face-centered cubic complex concentrated solid-solution alloys (CSA) have been evaluated for phase stability, mechanical properties, and radiation damage effects from heavy ions. Cr18Fe27Mn27Ni28 and Cr15Fe35Mn15Ni35 were predicted by thermodynamic calculations to phase separate and maintain a single phase at 700 & nbsp;C, respectively. Aging experiments at this temperature confirmed varying degrees of precipitation of a body-centered cubic phase in both Cr18Fe27Mn27Ni28 and Cr15Fe35Mn15Ni35. The alloys showed promising strength in tensile deformation at room temperature, with yield strengths of 155 MPa and 151 MPa for Cr18Fe27Mn27Ni28 and Cr15Fe35Mn15Ni35, respectively. At 500 & nbsp;C, the yield strength of Cr18Fe27Mn27Ni28 fell to 93 MPa, and to 100 MPa in Cr15Fe35Mn15Ni35. Unlike Cr18Fe27Mn27Ni28, Cr15Fe35Mn15Ni35 gained some ductility at 500 C compared to room temperature. The two CSAs were irradiated to 75 dpa at 500 C in the plateau region of the displacement curve using 3.7 MeV Ni2+ ions, alongside model alloy 709 as a reference. Irradiation results produced similar densities and sizes of dislocations loops in the two CSAs compared to the reference. However, while large voids form in the plateau region of Cr18Fe27Mn27Ni28, small voids form just beyond the displacement peak of Cr15Fe35Mn15Ni35. Atom probe tomography and energy dispersive X-ray spectroscopy-equipped scanning transmission electron microscopes were used to characterize the alloys for changes in chemical distribution. (C)& nbsp;2022 Elsevier B.V. All rights reserved.
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