基于NOR Flash实现KV数据库的应用研究

为在水文自动测报系统遥测站的遥测终端机中基于NOR Flash存储器可靠地存放系统运行参数,分析遥测终端机中常用固态存储器的读写特点和KV键值数据库的实现方式,设计一种基于W25Q256存储器实现KV键值数据库的解决方案.方案充分应用NOR Flash存储器的读写特点,针对同一个地址重复写入的最小编程颗粒度为1 bit,结合KV键值实现一个键值对状态的顺序变化,进而通过在存储器的一段空间内顺序增加地址存放KV键值对,实现KV键值对的新增、修改、删除和检索操作,达到磨损均衡和异常掉电恢复效果.在嵌入式系统中,相较于常用的采用固定地址保存运行参数的方案,该研究方案能有效延长NOR Flash存储器的使用寿命,提高系统的稳定可靠性.     

祁连山地区钨矿特征及找矿前景分析

结合区带成矿地质背景、矿点地质特征、矿化类型,对区带内自钨矿成因类型进行了划分(即石英脉一矽卡岩型、矽卡岩型、石英脉型等三大类);依据成矿物质来源、成矿时代、控矿因素、成矿系列等,对其成矿规律进行了总结;在此基础上进一步分析了区带的找矿前景。     

Studies on immobilization of thorium in barium borosilicate glass

The barium borosilicate glass (BBS) matrix has shown considerable solubility of ThO₂ at 1000 °C. As seen by X-ray diffractometry (XRD) and Electron probe micro analysis (EPMA) up to 15.86 wt% of ThO₂ could be dissolved in this matrix. The homogeneity of thoria loaded glass was convincingly ascertained by EPMA. Attempts to load more than 16 wt% ThO₂ led to the phase separation of crystalline phases identified as major phase of ThO₂ and minor percentage of ThSiO₄ phase with altogether different morphologies, as seen by XRD. Interestingly, the back scattered images of thorite crystals point towards the presence of chemical zoning. The results being reported in this paper are of interest especially with respect to immobilization of other actinides in borosilicate glass matrix.     

Role of microstructure and heat treatments on the desorption kinetics of tritium from austenitic stainless steels

The liquid scintillation counting of solid samples (LSC-SS technique) was successfully used to study the role of microstructure and heat treatments on the behavior of residual tritium in several austenitic stainless steels (as-cast remelted tritiated waste, 316LN and 321 steels). The role of desorption annealing in the 100-600 °C range on the residual amount of tritium in tritiated waste was investigated. The residual tritium concentration computed from surface activity measurements is in good agreement with experimental values measured by liquid scintillation counting after full dissolution of the samples. The kinetics of tritium desorption recorded with the LSC-SS technique shows a significant desorption of residual tritium at room temperature, a strong barrier effect of thermal oxide films on the tritium desorption and a dependance of the tritium release on the steels microstructure. Annealing in the 300-600 °C range allows to desorb a large fraction of the residual tritium. However a significant trapping of tritium is evidenced. The influence of trapping phenomena on the concentration of residual tritium and on its dependance with the annealing temperature was investigated with different recrystallized and sensitized microstructures. Trapping is evidenced mainly below 150 °C and concerns a small fraction of the total amount of tritium introduced in austenitic steels. It presumably occurs preferentially on precipitates such as Ti(CN) or on intermetallic phases.     

A glass-encapsulated calcium phosphate wasteform for the immobilization of actinide-, fluoride-, and chloride-containing radioactive wastes from the pyrochemical reprocessing of plutonium metal

Chloride-containing radioactive wastes are generated during the pyrochemical reprocessing of Pu metal. Immobilization of these wastes in borosilicate glass or Synroc-type ceramics is not feasible due to the very low solubility of chlorides in these hosts. Alternative candidates have therefore been sought including phosphate-based glasses, crystalline ceramics and hybrid glass/ceramic systems. These studies have shown that high losses of chloride or evolution of chlorine gas from the melt make vitrification an unacceptable solution unless suitable off-gas treatment facilities capable of dealing with these corrosive by-products are available. On the other hand, both sodium aluminosilicate and calcium phosphate ceramics are capable of retaining chloride in stable mineral phases, which include sodalite, Na₈(AlSiO₄)₆Cl₂, chlorapatite, Ca₅(PO₄)₃Cl, and spodiosite, Ca₂(PO₄)Cl. The immobilization process developed in this study involves a solid state process in which waste and precursor powders are mixed and reacted in air at temperatures in the range 700-800 °C. The ceramic products are non-hygroscopic free-flowing powders that only require encapsulation in a relatively low melting temperature phosphate-based glass to produce a monolithic wasteform suitable for storage and ultimate disposal.     

Processing glass-pyrochlore composites for nuclear waste encapsulation

Glass matrix composites have been developed as alternative materials to immobilize nuclear solid waste, in particular actinides. These composites are made of soda borosilicate glass matrix, into which particles of lanthanum zirconate pyrochlore are encapsulated in concentrations of 30 vol.%. The fabrication process involves powder mixing followed by hot-pressing. At the relatively low processing temperature used (620 °C), the pyrochlore crystalline structure of the zirconate, which is relevant for containment of radioactive nuclei, remains unaltered. The microstructure of the composites exhibits a homogeneous distribution of isolated pyrochlore particles in the glass matrix and strong bonding at the matrix-particle interfaces. Hot-pressing was found to lead to high densification (95% th.d.) of the composite. The materials are characterized by relatively high elastic modulus, flexural strength, hardness and fracture toughness. A numerical approach using a microstructure-based finite element solver was used in order to investigate the mechanical properties of the composites.     

Leaching behaviour of unirradiated high temperature reactor (HTR) UO2-ThO2 mixed oxides fuel particles

The dissolution of different mixed oxide (U, Th)O₂ particles under reducing conditions has been studied using a continuous flow-through reactor. The U/Th ratio seems to have no or little influence on the normalised leaching rate of thorium or uranium, The release rate of uranium from the outer surface of a Th rich matrix seems to follow the behaviour of pure UO₂ even though U is a minor component in these phases and the dissolution rate of Th is much lower. After long time U concentrations will become depleted at the solids surface and it will be expected that U release rates will become controlled by the release rates of thorium (rates at neutral pH < 10⁻⁶ g m⁻² d⁻¹). Under reducing conditions, the matrix of HTR fuel particles presents significant intrinsic radionuclide confinement properties.     

Radiolytic modelling of spent fuel oxidative dissolution mechanism. Calibration against UO2 dynamic leaching experiments

Calibration and testing are inherent aspects of any modelling exercise and consequently they are key issues in developing a model for the oxidative dissolution of spent fuel. In the present work we present the outcome of the calibration process for the kinetic constants of a UO₂ oxidative dissolution mechanism developed for using in a radiolytic model. Experimental data obtained in dynamic leaching experiments of unirradiated UO₂ has been used for this purpose. The iterative calibration process has provided some insight into the detailed mechanism taking place in the alteration of UO₂, particularly the role of OH radicals and their interaction with the carbonate system. The results show that, although more simulations are needed for testing in different experimental systems, the calibrated oxidative dissolution mechanism could be included in radiolytic models to gain confidence in the prediction of the long-term alteration rate of the spent fuel under repository conditions.     

高温AlGaInP/GaAs双异质结双极晶体管

利用Mo/W/Ti/Au难熔金属作为发射极欧姆接触设计并制作了一种用于功率放大器的新结构AlGaInP/GaAs双异质结双极晶体管（DHBT），分析了与传统AuGeNi作为接触电极的AlGaInP/GaAs DHBT的直流特性差异．研究结果表明，利用难熔金属作为欧姆接触电极的DHBT器件具有较好的高温特性，并进一步分析了其具有良好高温特性的机理．     

Thermal analysis and tests of W/Cu brazing for primary collimator scraper in CSNS/RCS

To the transverse beam collimation system in a rapid cycling synchrotron,an important component is the primary collimator,which improves emittance of the beam halo particles such that the particles outside the predefined trajectory can be absorbed by the secondary collimators.Given the material properties and power deposition distribution,the beam scraper of the primary collimator is a0.17 mm tungsten foil on a double face-wedged copper block of 121.5 mm x 20 mm.The heat is transferred to the outside by a φ34 mm copper rod.In this paper,for minimizing brazing thermal stress,we report our theoretical analysis and tests on brazing the tungsten and copper materials which differ greatly in size.We show that the thermal stress effect can be controlled effectively by creating stress relief grooves on the copper block and inserting a tungsten transition layer into the copper block.This innovation contributes to the successful RD of the primary collimator.And this study may be of help for working out a brazing plan of similar structures.