用于铀微粒年龄测量的铀钍氧化物混合微粒SIMS测量研究

放射性材料的年龄信息是一项重要的溯源指纹特征,铀微粒年龄测量研究对于核取证技术应用具有重要意义。本工作通过使用二次离子质谱（SIMS）、电感耦合等离子体质谱（ICP-MS）测量自制单分散铀钍氧化物混合微粒获得了单个微米级微粒中铀钍比值的相对灵敏度因子（RSF_Th/U）,结合扫描电子显微镜（SEM）等常规分析技术,确定了最佳测量条件,探索了微米级铀钍混合微粒的SIMS测量方法。测量结果表明,对于粒径为2~3 μm的混合微粒,不同微粒间²³²Th/²³⁸U比值的相对标准偏差小于3%(n=12),平均RSF_Th/U为1.259±0.032。通过测量年龄已知的铀同位素固体标准物质CRM970对RSF_Th/U进行了验证。结果表明,对于粒径为5~10 μm的CRM970铀粉末样品,年龄测量结果准确,相对标准偏差为3%(n=16)。该方法受干扰信号影响较小,测量结果稳定,可用于微米级铀微粒年龄的测量。     

Recognition of uranium oxides in soil particulate matter by means of μ-Raman spectrometry

Soil samples from an abandoned uranium mine have been investigated in order to determine the molecular phases of uranium compounds. The experiments were carried out with soil particulate matter, collected randomly from the area of the formerly exploited ore. To select the particles rich with uranium, scanning electron microscopy with energy-dispersive X-ray attachment (SEM/EDX) was applied first. Afterwards, the particles were relocated and measured by μ-Raman spectrometry (MRS). Residues of the main deposit, uraninite UO₂, were detected, along with its alteration products. In terms of Raman scattering properties, uranium oxides are quite sensitive to the laser beam wavelength, which results in very specific features of their Raman spectra. In this paper the Raman spectra of uranium oxides of different origin and oxidation states, measured with 514 and 785 nm lasers, are also presented.     

New fuel rod design based on LEU/Th fuel and an assessment on in-core fuel cycle length of the MHR core

The current Modular Helium Reactor (MHR) fuel cycle uses fissile LEU (19.8 wt% U-235) and depleted uranium in separate TRISO particles, in a single fuel rod within a graphite matrix. The TRISO particle volume packing fraction (PF) in the fuel rods is 29%, of which the LEU particle PF is 62%. The lifetime between refuelings is about 476 effective full power days (EFPD). In this paper we assess the possibility of replacing the depleted uranium TRISO particles with thorium TRISO particles, and evaluate the impact of such replacement on fuel cycle length. A preliminary scoping study was performed to determine the most promising fuel rod/zoning configurations. The scoping study indicates that there is advantage to separating the thorium TRISO particles from the LEU particles at the fuel rod level instead of mixing them within a single rod. An axial checkerboard distribution of the fuel rods where all uranium and all thorium rods are interchangeable along the axial direction within the graphite block is the most promising configuration that was identified in this study and can be lead to a fuel cycle length extension of 50-80% relative to the current design, with only a modest increase in the fissile material loading (15-20%). To this advantage can be added the benefit of a significant reduction in nuclear waste and in health risk. This study also lays the foundation for improving the fuel rod arrangement within the graphite block and the graphite blocks within the entire reactor core. The analysis is limited to a once - through fuel cycle based on in situ fissioning of the U-233, without further separation and reprocessing. The preliminary heat transfer analysis indicates that the maximum temperature in the fuel will be raised by about 10-15% over that of current MHR design.     

基于热电离质谱的单微粒铀同位素分析方法比较

在核保障中,测定擦拭样品单微粒中铀同位素比可为探测未申报或隐匿的核活动提供重要的信息。裂变径迹 热电离质谱（FT TIMS）是该领域中的主流技术之一,近年来提出了一种新技术——扫描电子显微镜 热电离质谱（SEM TIMS）。本文通过对标准微粒样品的测定系统评估比较了该两种方法。该两种方法的测量结果均与标准微粒的参考值吻合较好,FT TIMS易于寻找、定位易裂变核素,对高浓铀具有极高的灵敏度。SEM TIMS则更易于同步开展含铀微粒形貌学和元素组成等分析,可视作FT TIMS有益的补充。     

草酸铈和草酸铀的热分解机理及煅烧遗传性北大核心CSCD

在核燃料后处理过程钚尾端处理中,通常利用草酸制备草酸钚沉淀然后进一步煅烧获得氧化钚颗粒、用于MOX燃料的制造,因此掌握草酸钚热分解机制、控制氧化钚产品颗粒的形貌和粒度具有实际意义。针对草酸盐的热分解机理和沉淀在煅烧后颗粒形貌遗传性问题,本工作选用草酸铈和草酸铀作为研究对象,系统研究了草酸铈、草酸铀的热分解反应,结合同步热分析仪(TG/DSC)与X射线衍射仪(XRD)的表征,获得了草酸铈和草酸铀的热分解数据;制备了不同粒度的草酸铈和草酸铀,用激光粒度仪考察了煅烧前后颗粒的粒度,并用扫描电镜(SEM)观察颗粒聚集状态、粒度和形貌,结果表明煅烧分解会导致粒径有规律地下降,但形貌得以保留。     

磁控溅射制备金属铀膜

研究了通过磁控溅射方法制备高纯金属铀膜的可行性。采用X射线衍射（XRD）、俄歇电子能谱（AES）、扫描电镜（SEM）、表面轮廓仪分析了沉积在单晶硅或金基材上铀薄膜的微观结构、成分、界面结构及厚度、表面形貌和表面粗糙度。分析结果表明：磁控溅射制备的铀薄膜为纯金属态,氧含量和其它杂质含量均低于俄歇电子能谱仪的探测下限;溅射沉积的铀镀层与铝镀层之间存在界面作用,两者相互扩散并形成合金相,扩散层厚度约为10nm。铀薄膜厚度可达微米级,表面光洁,均方根（RMS）粗糙度优于15nm。     

抽气碰撞法回收擦拭样品上铀微粒方法

建立了一种回收微粒的新方法——抽气碰撞法。擦拭布上的铀微粒通过抽气碰撞装置回收到导电胶上,用于二次离子质谱仪（SIMS）对微粒的同位素丰度比测量。使用扫描电镜（SEM）寻找和统计擦拭布和导电胶上的微粒数目,计算装置的回收率。该装置对核孔膜上直径为0.5～20.0μm的铅微粒回收率为(43±5)%,擦拭布上铀微粒回收率约为48%,回收微粒的分散性好。制备的样品可直接用于SIMS测量,SIMS对²³⁵U与²³⁸U同位素丰度比的测量值为0.00725±0.00009,测量标准偏差小于3%。     

磁性纳米微粒的制备方法及其在放射免疫分析中的初步应用

采用化学共沉淀法合成水溶性正癸酸包覆的Fe3O4磁性纳米微粒,并以此为核心物理吸附羊抗兔IgG制备磁性纳米第二抗体,作为磁性分离载体用于放射免疫分析中。用光子相关光谱仪与透射电镜测定了磁性纳米微粒的粒径大小、粒径分布和多分散度等。研究了羊抗兔IgG包被磁性纳米微粒的包被介质的pH、羊抗兔IgG用量和包被时间等制备条件对磁性纳米微粒二抗免疫反应结合能力和非特异结合的影响。结果表明：化学共沉淀法制得的Fe3O4平均粒径约为10-20nm,吸附羊抗兔IgG后其平均粒径为1000nm左右,并可用于放射免疫分析中。提示羊抗兔IgG可通过物理吸附的方法固定在磁性纳米微粒上,其优点是制备方法简便、省时和成本低,在放射免疫分析中具有磁性分离的方便性,具有较大的医学应用价值。     

Use of Neural Network for the Simulation of a Gas Centrifuge

The prediction by a mathematical model of the separation of uranium isotopes using a gas centrifuge process is a hard task. The gas motion can be described by analytical or numerical solutions of the system of equations defined by the equation of continuity, the Navier-Stokes equation and the equation of energy. However, these calculations cannot be performed for actual centrifuges.

Neural networks are an alternative for modelling complex problems that show too many difficulties to be solved by phenomenological models.

The authors propose the use of neural networks for the simulation and prevision of the separative and operational parameters of a gas centrifuge separating uranium isotopes. The results from the uranium separation experiments (Zippe data) are compiled and presented to the neural network in the learning and testing processes. The prediction using the neural network model shows good agreement with the experimental data.     

活性炭吸附法测量铀尾矿氡析出率

介绍一种用活性炭吸附法测量铀尾矿表面氡析出率的方法，简要叙述了采样器和测量仪器的性能、特点，刻度及测量方法。通过在铀尾矿库的现场连续一年的监测，研究铀尾矿库年氡析出率的变化规律，总结测量年平均氡析出率的最佳季节和条件。     

胺肟基尼龙66纤维的制备及其对模拟核工业废水中铀的选择性吸附

结合辐射接枝聚合和化学修饰,通过三步处理,成功地制备了偕胺肟化尼龙66纤维。通过傅里叶红外光谱(Fourier transform infrared spectroscopy,FTIR)仪和扫描电子显微镜(Scanning electron microscopy,SEM)对初始的和修饰后的尼龙66纤维进行化学结构和微观形貌表征。结果表明,偕胺肟基团以共价方式连接到尼龙66纤维上。通过模拟核工业废水的吸附研究表明,PA66-g-PGMA-IDPAO纤维对铀酰离子有高的吸附效率(91.3%)和吸附选择性,具有巨大的潜在工业应用价值。     

An electromagnetic apparatus with a high resolving power for the separation of the isotopes of the heavy elements

The operating principles and the main features of an electromagnetic apparatus which has been designed and constructed for the separation of isotopes with a relative mass difference of about 1/240 are here described. An axially-symmetric magnetic field is used (focussing angle 225 °) in which the dispersion is proportional to the square of the focussing angle. The dispersion obtained with the separation apparatus is 20 mm for a 1% relative difference in mass. The magnetic field of the apparatus is stabilized to within 0.005% with the help of a circuit containing electron-beam regulators. Ion sources, with a discharge in the vapor of the working and auxiliary substances, were developed for the separation of large and small quantities of the raw material. General-purpose ion collectors were also developed. Some results of the separation of lead, uranium and plutonium isotopes are given. The enrichment factor in the separation of lead isotopes reaches 300, while for uranium and plutonium it is ～ 1000.     

Method for Calculating and Optimizing a Cascade for Separating a Mixture of Isotopes with Impurities

A method is proposed for calculating and optimizing a cascade separating a mixture of isotopes with impurities. The approach developed optimizes a cascade with arbitrary mixture composition and a large difference in the masses of the components being separated. The application of this method to the separation of a binary mixture of uranium isotopes with impurities is examined. The properties of cascades optimized with respect to the criterion of minimum total flux are investigated. Their correspondence to R cascades with symmetric steps is analyzed. A comparison is made with optimal cascades with no impurities. It is shown that multicomponent potentials can be used to estimate the separation efficiency for a binary mixture of isotopes with an impurity.__________Translated from Atomnaya Energiya, Vol. 98, No. 4, pp. 293–300, April, 2005.     

Industrial methods of processing poor uranium ores

In recent years great progress has been achieved in the technology of processing uranium ores. A method of extracting uranium from solutions by means of absorption on resins, developed comparatively recently, now makes it possible to obtain more than 70% of the total extractable uranium. The extraction of uranium with liquid extractants also finds wide application, especially in plants using acid leaching of uranium.The normal methods of mechanical enrichment (gravitation, flotation, etc.) play a comparatively small part. However, with complex and poor ores being used, mechanical enrichment in the processing of uranium ores assumes greater importance. Radiometric enrichment is particularly progressive and this makes use of the radioactivity of uranium minerals for separating them from barren rock.     

Numerical Solution Method for Prediction of Heat Transfer in Tube Bundles of Large-Size Intermediate Heat Exchangers in FBR Plants

A study was made of uranium contamination in (a) the coating layers of TRISO particles (a-1) before compacting and (a-2) separated from once-compacted fuel heat-treated at 1,400 or 1,800°C, and (b) in the matrix material of the same compacts. The contamination in the pyrocarbon layers of the coating was determined, after mechanically separating the coating layers, by a procedure of neutron activation, burn-off and ¹³³Xe trapping. For the silicon carbide coating layer, the fragments of coating left from the above procedure were fused into alkaline melt, and the ¹³³Xe released at each heating step was trapped. For the matrix material, the fuel compacts were deconsolidated electrolytically or mechanically, followed by activation analysis. The results of the foregoing measurements proved the uranium contamination in pyrocarbon and silicon carbide coating to be at most of the order of 10^?4 in reference to uranium content in kernel, while the corresponding value for particles sampled from fuel compacts heat-treated at 1,800°C were appreciably higher. The corresponding values found for the matrix material were of the order of 10^?5.     

冠醚包覆聚苯乙烯磁性颗粒的合成、表征及其对锶离子的吸附性能

为建立应用磁性颗粒快速分离低水平~(90)Sr的方法,分别以沉淀法合成Fe_3O_4纳米颗粒,微乳聚合法合成磁性聚苯乙烯-二乙烯苯颗粒(Pst-DVB@Fe_3O_4),并将4,4’(5’)-二叔丁基二环己基-18-冠-6(DtBuCH18C6)通过物理作用包覆在Pst-DVB@Fe_3O_4颗粒表面,最终生成一种新型的核-壳结构的DtBuCH18C6包覆的聚合物磁性颗粒(DtBuCH18C6@Pst-DVB@Fe_3O_4)。采用扫描电镜、红外光谱、X射线光电子能谱等方法对颗粒的形貌、结构、表面元素组成及含量进行分析。用于分离水溶液中Sr~(2+)的批实验结果表明,Sr~(2+)的吸附量随着pH值而增加,而在pH=11和pH=13之间急剧上升,温度升高,其吸附量增加。吸附动力学模型结果表明,Sr~(2+)在DtBuCH18C6@Pst-DVB@Fe_3O_4表面的吸附更接近于其和Sr~(2+)之间形成络合物的单层化学吸附过程,符合Langmuir模型。在优化的条件下其最大吸附容量为2.62 mg/g。     

铀烧绿石基玻璃陶瓷固化体的合成及化学稳定性评估

为改善核素铀在玻璃陶瓷固化体中包容量低、亲玻璃而疏陶瓷的赋存问题，本文采用预处理与熔融-热处理相结合的方法，制备了含铀母玻璃（PG）和铀烧绿石基玻璃陶瓷（GC）固化体，并借助X射线衍射（XRD）、扫描电镜（SEM-EDS)、透射电子显微镜（TEM）、X射线光电子能谱（XPS）等检测手段，表征了GC固化体的物相结构并评估了其化学稳定性。XRD结果表明，GC固化体中的铀烧绿石是由PG中预先生成的萤石晶核发生相变而形成的，且是GC固化体中的主晶相；SEM和TEM结果显示，铀烧绿石相在玻璃基体上主要呈四方形均匀生长，且与玻璃的相容性好；元素分析结果证实，GC固化体中的铀高度富集于烧绿石中，残留于玻璃中的量极少；由EDS推算出铀烧绿石的化学计量式为(Ca_1.05Na_0.20U_0.73)(Ti_1.48Al_0.58)O_7+x，其固溶量与设计值相当，实现了铀在陶瓷相中的较大固溶；XPS证实PG和GC固化体中的铀均以+4价居多，这为烧绿石包容更多铀提供了价态上的可行性；MCC-1结果表明GC固化体抗水性良好。     

Plasma separation of the elements applied to nuclear materials handling

Physical plasma method of separation by mass of the constituent elements together with chemical methods could find application in the fuel cycle of fast recators. These methods could also be useful for separating decay products during reprocessing of spent fuel of existing reactors and in the solution of questions concerning safety and nonproliferation. A possible plasma separation scheme is presented: solid-state material is transferred in a magnetic field into a low-temperature plasma flow and ions in a prescribed mass range are selectively accelerated and then spatially separated in a toroidal magnetic field. __________ Translated from Atomnaya énergiya, Vol. 101, No. 4, pp. 302–306, October, 2006.     

TEVA-TEVA色层分离大量铀中的微量镎

建立了高效分离高铀镎比样品中微量镎的方法。采用TEVA-TEVA萃取色层柱分离铀产品中微量镎，经3.0 mol/L HNO₃淋洗过后以0.5 mol/L (NH₄)₂CO₃洗脱TEVA树脂上的镎，加入HNO₃分解洗脱液中的碳酸镎酰离子并将溶液调至3.0 mol/L HNO₃，进行TEVA柱二次分离，最后采用0.02 mol/L HNO₃-0.02 mol/L HF洗脱。结果表明：该法稳定性较好，测量相对标准偏差优于4%（n=3），镎的平均回收率大于86%，铀的去污因子大于10⁵，适用于高铀镎比样品中微量镎的分离，在乏燃料后处理工艺中铀线尾端的微量镎分析中具有较好的应用前景。     

Molecular laser isotope separation versus atomic vapor laser isotope separation

In this work, the competitive molecular and atomic laser isotope separation methods are investigated to evaluate the potential of each une for more economical fuel fabrication of enriched uranium celas which can be practically used in light water reactors (LWR). The advantages, drawbacks and feasibility of various techniques of laser isotope separation including AVLIS and the main MLIS methods such as MOLIS, CRISLA, and SILARC techniques have been reviewed. Laser isotope separation (LIS) deserves an intensive research, because laser techniques have a number of advantages over traditional ones, such as highly selective elementary separation event, possibility of separating the required isotope, low energy consumption, short start up time and practically single-stage production, which may economize fuel processing [11.