Beryllium oxide (BeO) is a promising material in the nuclear field. However, preparing BeO ceramics is challenging because pure BeO powders are hard to sinter densely. Previous studies have mainly focused on the aspects of microstructure such as grain size, porosity, and macro-thermodynamic properties of sintered BeO ceramics. However, such studies provide limited insight into the atomic-scale sintering mechanism. Therefore, in this study, molecular dynamics (MD) simulations were performed to investigate the sintering behavior of BeO powders, with Aluminium Oxide (Al2O3) and Silicon Oxide (SiO2) as sintering aids. The results demonstrate that Al2O3 exhibits the most significant improvement in the diffusivity of the system, but the system with only Al2O3 as the sintering aid shows weaker kinetic stability than the composite-aid systems. Thus, composite sintering aids are a more optimal choice. Specifically, the sintering aid with an Al2O3: SiO2 ratio of 1:3 is most effective in promoting the densification of BeO during sintering. These findings should provide crucial insights to further understand the sintering mechanism, which is beneficial for developing high-performance BeO ceramics.

The irradiation behavior of beryllium oxide (BeO) provides key information for the design of novel nuclear reactor. The post-irradiation annealing microbehavior of BeO was studied via molecular dynamics (MD) simulation method and compared with previous literature results. It was found that during annealing process, the irradiated BeO would have amorphous-to-crystallization phase transition. Moreover, β -BeO was appeared during annealing process and 10% (mass percent) could be reached. It was the first time to report the phenomenon of amorphous to β -BeO transformation during BeO annealing, and the mechanism of the transformation was discussed as well. Simulation results showed that the annealing process was also a process of reducing internal stress and potential energy of the system. The MD simulation results were of great significance for the understanding the evolution behavior of BeO materials after irradiation, which presented a foundational reference for the materials irradiation annealing with MD in the future.

In recent years, beryllium oxide (BeO) ceramics have been reconsidered for application in micro nuclear reactors. The applications of BeO ceramics in micro nuclear reactors are mainly as neutron moderator, neutron reflector material or nuclear fuel pellet matrix material. When used in these reactors, BeO is often subjected to extreme environments. Therefore, it is necessary to master the high temperature mechanical properties of BeO. In this paper, the mechanical properties of BeO ceramics at high temperature were studied experimentally, which shows a trend that first increasing, then decreasing and at last increasing again. Besides, the corresponding mechanism was discussed and it was concluded that the glass phase between the grains of BeO is the key factor affecting its high temperature strength. The mechanism was verified by molecular dynamics (MD) simulation. This paper is of great significance for the application of BeO ceramics.

In this research paper, two types of tristructural isotropic (TRISO) particles were dispersed within beryllium oxide (BeO) matrix using the pressureless sintering method to create fully ceramic microencapsulated (FCM) nuclear fuel. The findings indicated that TRISO particles with an outermost layer comprising silicon carbide (SiC) can establish interfacial bonding with BeO matrix, while those with an outermost layer comprising pyrocarbon (OPyC) and BeO matrix exhibit a noticeable gap after the sintering process. Results from transmission electron microscopy (TEM) revealed the presence of an amorphous phase transition zone between SiC and BeO. This zone, according to molecular dynamics (MD) simulation results, is an amorphous phase composed of Be, Si, and O atoms. Overall, TRISO particles with an outermost layer of SiC are more suitable for sintering with BeO matrix than those with an outermost layer of OPyC.

Due to its excellent strength and high thermal conductivity, beryllium oxide (BeO) is widely used as a fuel matrix material and neutron moderator in nuclear reactors. In the preparation of bulk BeO materials, previous research on the hot-pressing sintering methods is insufficient. This paper mainly introduces the process of hot-pressing sintering, and uses orthogonal experiments with different process parameters to explore their influences on the grain size and porosity of the products, and how to affect their flexural strength and thermal conductivity as well. Molecular dynamics simulations have also been used to study the growth of the amorphous phase in BeO ceramics. In addition, a relatively optimized preparation scheme, that is, holding at 1500 degrees C for 30 min and pressure of 20 MPa can be obtained. Under the optimized parameter, the flexural strength of the product can reach 244.74 MPa, and its thermal conductivity can reach 179 W/(m*K) at 200 degrees C, and 31 W/(m*K) at about 1000 degrees C. It provides a reference for the preparation of high performance BeO ceramics.

Beryllium oxide (BeO) has recently regained attention as a material in nuclear reactors. However, as an oxide ceramic, it is susceptible to the effects of a water environment, which can impact its service performance. This paper presents an experiment on the stress corrosion behavior of BeO in a water environment at room temperature (RT), along with a discussion of the relevant mechanisms. The study first conducted flexural strength tests in three different environments: vacuum, air, and water. It then performed Vickers hardness tests in both air and water environments. The results of both tests indicated a significant decrease in the strength of BeO in water environment, even at RT. Furthermore, the use of Time-of-Flight Secondary Ion Mass Spectrometry (TOF-SIMS) and molecular dynamics simulations revealed that the stress corrosion of BeO at RT was due to the breakage of the original Be-O bond, followed by the formation of a new Be-O bond with an O atom in H2O. This resulted in a reduction in the driving force needed for crack initiation and propagation. The findings of this study have practical implications for further research into the stress corrosion phenomenon of BeO at higher temperatures.

Beryllium oxide (BeO) is being re-emphasized and utilized in Micro Modular Reactors (MMR) because of its prominent nuclear and high temperature properties in recent years. The implications of the research about effects of neutron irradiation on the microstructure and properties of BeO are significant. This article comprehensively reviews the effects of neutron irradiation on BeO and proposes the maximum permissible neutron doses at different temperatures for BeO without cracks in appearance according to the data in the previous literature. This maximum permissible neutron dose value has important reference significance for the experimental study of BeO. The effects of neutron irradiation on the thermal conductivity and flexural strength of BeO are also discussed. In addition, microstructure evolution of irradiated BeO during post-irradiation annealing is summarized. This review article has important implications for the application of BeO in MMR.

This article investigated the deformation behavior of nano-polycrystalline beryllium oxide under tensile and compressive stress using the molecular dynamics simulation method. Both the tensile and compressive test results indicate that beryllium oxide breaks mainly along grain boundaries. At low temperature, there is little internal deformation of beryllium oxide grains. When the temperature is above 1473 K, the internal deformation of beryllium oxide grains also occurs, and the phenomenon becomes more obvious with the increase in temperature. This deformation within the grain should be plastic. The flexural strength fracture morphology of beryllium oxide also shows that the fracture mode of beryllium oxide is a brittle fracture at low temperature, while the slip bands appear at 1773 K. This indicates that beryllium oxide, as a ceramic material, can also undergo plastic deformation under high temperature and stress.

Beryllium oxide as a nuclear material has received renewed attention in recent years, and it is critical to investigate the irradiation behavior. The irradiation damage accumulation in beryllium oxide is investigated by molecular dynamics method. The evolution of defects and the degree of amorphization have been revealed as a function of dose at different irradiation temperatures. The dislocation induced by irradiation is obvious only when the irradiation temperature is above 873 K. A new cubic diamond structure appears at higher irradiation temperatures. Simulated XRD results are closely related to the degree of amorphization. In addition, the influence of irradiation damage on the compressive strength of beryllium oxide under different irradiation conditions is also studied.

In recent years, beryllium oxide has been widely utilized in multiple compact nuclear reactors as the neutron moderator, the neutron reflector or the matrix material with dispersed nuclear fuels due to its prominent properties. In the past 70 years, beryllium oxide has been studied extensively, but rarely been systematically organized. This article provides a systematic review of the application history, thermal properties, mechanical properties, corrosion behavior and fabrication methods of beryllium oxide. Data from previous literature are extracted and sorted out, and all of these original data are attached as the supplementary material, so that subsequent researchers can utilize this paper as a database for beryllium oxide research in reactor design or simulation analysis, etc. In addition, this review article also attempts to point out the insufficiency of research on beryllium oxide, and the possible key research areas about beryllium oxide in the future.

In the present study, the microstructure, toughness and cracking behavior of a batch of X12CrMoWNbVN10-1-1 martensitic heat resistant steels used for manufacturing steam turbine blades in power plant were investigated. The manufacturing route of the X12CrMoWNbVN10-1-1 martensitic steels was casting, forging and heat treatment. These martensitic steels were taken from different positions in the same ingot, but they exhibited different toughness, which were marked as two groups: (i) the matrix located in the core of the ingot, hereinafter referred to sample A; (ii) the matrix located in the edge of the ingot, hereinafter referred to sample B. Impact toughness was characterized by Instrumented Charpy impact test, fracture toughness was determined by single edge notch tension (SENT) test in terms of double-clip gauge method, and crack propagation behavior was observed by in-situ single edge notch bend (SENB) test. Optical microscopy (OM), scanning electron microscopy (SEM) and high-resolution energy dispersive X-ray spectrometry (EDX) were used to examine the microstructure and cracking features. The impact toughness of sample A was much lower than that of sample B, which was proved to be attributed to the large amount of NbC precipitation in sample A. The clustered NbC particles could induce brittle cracking and thus significantly reduced absorbed energy for stable crack propagation, resulting in the poor impact toughness of sample A. It was believed that the large amount of NbC in sample A was precipitated directly from liquid metal during casting process, which might be related to the high niobium content of the core of the ingot due to the low cooling rate. The results revealed that the large number of clustered NbC particles formed in the casting process not only cannot promote precipitation-induced reinforcement, instead might be detrimental to toughness.

Based on Micro-Raman spectroscopy with the laser diameter of 400 nm, effects of different distribution of residual stresses in the depth direction on hardness and cutting performance of thin TiAlN coatings (about 4 mu m) on WC-10wt%Co substrate are studied. Residual stresses distribution maps from substrate to surface of two samples are acquired with the step length of 0.4 mu m. The hardness, modulus and H3/E2 respectively increase about 13.97%, 17.63% and 7.10% from Sample 1 to Sample 2. The milling experiments indicate that Sample 2 with higher compressive residual stress possesses better wear capacity and lower cutting forces than Sample 1. The reason is discussed: the deeper the crack propagates in the coating, the greater the residual compressive stress, which forces the crack to propagate along the direction of stress rather than along the cleavage plane, thus inhabiting the crack propagation along the vertical direction.

Geant4 Monte Carlo simulation results of the single event upset (SEU) induced by protons with energy ranging from 0.3 MeV to 1 GeV are reported. The SEU cross section for planar and three-dimensional (3D) die-stacked SRAM are calculated. The results show that the SEU cross sections of the planar device and the 3D device are different from each other under low energy proton direct ionization mechanism, but almost the same for the high energy proton. Besides, the multi-bit upset (MBU) ratio and pattern are presented and analyzed. The results indicate that the MBU ratio of the 3D die-stacked device is higher than that of the planar device, and the MBU patterns are more complicated. Finally, the on-orbit upset rate for the 3D die-stacked device and the planar device are calculated by SPACE RADIATION software. The calculation results indicate that no matter what the orbital parameters and shielding conditions are, the on-orbit upset rate of planar device is higher than that of 3D die-stacked device.

To study the effects of pulsed magnetic fields of different intensities on the dislocation density, residual stress, and hardness of Cr4Mo4V steel, magnetic treatment is conducted at 0, 1.0, 1.3, 1.5, 2.0, and 2.5 T. The dislocation density and residual stress are measured using Electron Backscatter Diffraction (EBSD) and X-ray technique, respectively. The results reveal the dislocation density and compressive residual stress decrease at lower magnetic fields such as 1.0 T and 1.3 T, while they increase at higher magnetic fields such as 2.0 T and 2.5 T. The average value of kernel averaged misorientation (KAM) and compressive residual stress decrease about 10.4% and 15.8%, respectively, at 1.0 T, while they increase about 5.88% and 18.2%, respectively, at 2.5 T. The average value of hardness decreases about 3.5% at 1.0 T, from 817 HV to 787 HV. With the increments of intensities, the hardness of the treated samples increases. The hardness essentially remains unchanged at 2.0 T and 2.5 T. The reason for the dislocation motion under the action of pulsed magnetic fields is discussed.

经过曲折探索,1938年科学家们发现了原子核的裂变现象,从此人类找到了释放核能的钥匙,开启了核能利用的新时代. 核电是清洁的绿色能源 核电是清洁的能源,是一种几乎没有温室气体释放的低碳绿色能源.核电站不用化石燃料,也就没有碳和杂质硫、氮等元素,以及这些元素在化学反应中产生的氧化物的排放,因此,核电对保护环境具有重要的意义.

Heavy-ion flux is an important experimental parameter in the ground based single event tests. The flux impact on a single event effect in different memory devices is analyzed by using GEANT4 and TCAD simulation methods. The transient radial track profile depends not only on the linear energy transfer (LET) of the incident ion, but also on the mass and energy of the ion. For the ions with the energies at the Bragg peaks, the radial charge distribution is wider when the ion LET is larger. The results extracted from the GEANT4 and TCAD simulations, together with detailed analysis of the device structure, are presented to demonstrate phenomena observed in the flux related experiment. The analysis shows that the flux effect conclusions drawn from the experiment are intrinsically connected and all indicate the mechanism that the flux effect stems from multiple ion-induced pulses functioning together and relies exquisitely on the specific response of the device.

核电站的核岛与常规岛 核燃料在核反应堆中发生的链式裂变反应会产生大量的热能,使水汽化,水蒸气通过管路进入汽轮机,推动汽轮发电机发电,使机械能转化成电能.整个过程是先由核能转化为热能,热能转化为机械能,机械能再转化为电能.核电站大体上可分为两个部分:一部分是利用核能产生蒸汽的系统,称之为核岛,核岛中主要的设备包括产生热量的核反应堆、导出热量的回路系统和蒸汽发生器;另一部分是利用蒸汽发电的系统,称之为常规岛,包括汽轮发电机组及其相应附属设备,这部分与普通火电站差别不大.

核电发展的历史回顾 核能发电的历史可以追溯到20世纪50年代"二战"后的冷战时期.1951年,美国首先进行了核能发电的试验.但首先建成核电站的却是苏联.1954年6月27日苏联在莫斯科近郊建成世界上第一座试验核电站——奥布灵斯克核电站,使核能的和平利用成为现实,也是人类和平利用原子能的成功典范.这是一座石墨水冷反应堆,电功率为5 000千瓦.虽然这座核电站规模很小,但是它已被全世界公认为人类科学与技术发展过程中的标志性成就.

什么是裂变链式反应 一个铀-235原子核的裂变大约放出200兆电子伏特的能量.如果只裂变一次,这个能量是难以利用的.幸运的是每次裂变过程中都会放出一些中子,这些新产生的中子又会打中另一个铀-235原子核,引起新的裂变,接着又放出中子又引起裂变,裂变过程将自己持续地依次进行下去,并源源不断地释放出能量.这个过程类似于多米诺骨牌,只要轻轻碰倒第一枚骨牌,其余的骨牌就会产生连锁反应,依次倒下.我们把这样一个过程称为自持裂变链式反应,它是利用原子裂变产生能量的重要理论基础,为核能的利用提供了现实的可能.