阴离子交换法从高通量堆辐照二氧化钍中分离铀

目前世界上已探明的具有开采价值的钍资源几乎与铀相等。Th/U燃料循环的主要优点在于~(233)U,相对于U/Pu燃料循环中的~(239)Pu,有更高的中子产额。事实上只有用钍作为再生燃料,热堆才可能实现增殖。作为钍基核燃料利用的基础研究的一部分,我们制订了一个从辐照二氧化钍中分离~(233)U( ~(232)U)的阴离子交换程序。每个辐照样品压成小药丸状、重80 mg的核纯ThO_2。装入辐照管中,在国内热功率为12.5万千瓦的高通量工程试验堆的铍反射层中辐照。热中子通量为2×10~(14)中子/cm~2·s,快热中子比为1:1。热中子积分通量约为1×10~(20)—1×10~(21)中子/cm~2。照好后的样品移至水池冷却,数月后处理。二氧化钍用含NH_4F及AlCl_3的浓HCl加热迥流溶解,然后将料液调至8 mol/1 HCl     

(DMHMP+HTTA)协同萃取硝酸钍的机理研究

螯合萃取剂噻吩甲酰三氟丙酮(HTTA)和各种中性磷类萃取剂在不同酸性介质中协同萃取钍已有报道,但与甲基膦酸二(1-甲庚)酯(DMHMP)的协同萃取机理迄今未见。我们曾研究过DMHMP萃取硝酸钍的机理。本文研究DMHMP与HTTA的苯溶液从硝酸介质中对硝酸钍的协同萃取。     

用α能谱法测定样品中~(233)U和~(232)U的相对含量

钍是一种潜在的核燃料。据国际原子能机构估计,目前已探明有开采价值的钍资源大致和铀相等。钍的充分利用可以大大丰富核能资源,增加核燃料储备。 钍铀核燃料循环的一个主要技术问题是钍在反应堆中受中子照射后,除生成有用的核燃料~(233)U外,还会产生~(232)U,一般化学方法不能使它与~(233)U分开。由于~(232)U的衰变子体放出能量     

一种基于粗糙集理论的智能故障诊断新方法

论文针对基于规则的故障源分离与定位方法中的一个关键问题,即诊断规则的获取,利用粗糙集的基本原理构造出了一种用于规则提取的新方法,其中包括了用于对故障决策表,即故障字典,进行属性约简的改进算法和属性值的顺序约简算法。该方法能够迅速从故障字典中提取出诊断规则,并揭示出故障信息内在的冗余性。最后实例应用的结果表明了该方法的有效性,尤其是在不完全信息情况下的有效性。     

末端基为喹啉基的两个开链冠醚与铀(Ⅵ),钍(Ⅳ)配合物的合成与表征

合成了末端基为喹啉基的两个开链冠醚2,2' (乙撑二氧)二[(8 喹啉氧基甲基)苯](L1)和2,2' (乙撑二氧)二[(8 喹哪啶氧基甲基)苯](L2)与U(Ⅵ),Th(Ⅳ)的配合物;并使用1HNMR,IR,差热 热重分析,摩尔电导率和元素分析等方法对配合物进行了表征;确定U(Ⅵ)配合物的组成为[UO2LNO3]NO3·3H2O(L=L1,L2);Th(Ⅳ)配合物的组成为[ThL(NO3)2](NO3)2·nH2O(L=L1,n=1;L=L2,n=3)。     

饰面石材中放射性核素钍的分析方法研究

在对天然饰面石材进行分类时，存在一种临界现象。即不能明确判断其属于A类，还是B类；或是B类，还是C类。影响石材分类的主要核素有钍、镭、钾。章着重研究了放射性核素钍在临界饰面石材分类中的检测方法。试样中的钍经酸溶分解、过滤，氨水沉淀、离心分离后，用硝酸溶解沉淀，在盐析剂硝酸铝作用下，用N-235(三辛胺)从硝酸溶液中萃取钍络合物，使钍与其他杂质分离；用盐酸溶液反萃取钍，在掩蔽剂存在下，用偶氮胂Ⅲ显色，于波长665nm处测定钍络合物的吸光度，其线性回归方程为Y=0.0956 46.8544X,y=0.9986；最低检出限为0.01μg/g。研究表明，该方法准确地解决了处于临界值时饰面石材的分类。     

Development of separation technology for the removal of radium-223 from targeted thorium conjugate formulations. Part I: purification of decayed thorium-227 on cation exchange columns

Targeted thorium conjugates (TTCs) are being explored as a potential future platform for specific tumor targeting pharmaceuticals. In TTCs, the alpha emitting radionuclide thorium-227 (²²⁷Th) with a half-life of 18.697 d is labeled to targeting moieties, such as monoclonal antibodies (mAbs). The amount of daughter nuclide radium-223 (²²³Ra, t_1/2?=_?11.435 d) will increase during manufacture and distribution, and so a technology for purification is required to assure an acceptable level of ²²³Ra is administrated to the patient. Since ²²³Ra is the only progeny of ²²⁷Th with a long half-life (days), the progenies of ²²³Ra will have a very limited stay in the formulation once ²²³Ra is removed. The focus in this study has, therefore, been on the removal of ²²³Ra. In this study, the sorption and separation of ²²³Ra (radium(II)) and ²²⁷Th (thorium(IV)) on cation exchange columns has been evaluated as a purification method of decayed ²²⁷Th (i.e. prior to radiolabelling of a mAb and formation of TTC). The goal is to minimize the sorption of ²²⁷Th and maximize the sorption of ²²³Ra. Statistical experimental design with formulation and process parameters, including buffered formulations comprising citrate and acetate, at various concentrations and pH, presence of free radical scavenger and chelator, and resin amount have been evaluated for impact on the purification process. The studies have been interpreted by the aid of multivariate data analysis. The correlations between design of experimental variables and sorption are summarized by regression models. The predictive accuracy of radionuclide sorption was given by standard deviation and 95% confidence intervals originating from statistical cross validation. Experimental results and statistical models for citrate-buffered formulations verified reproducible and acceptable sorption levels of ²²³Ra and ²²⁷Th under selected conditions. For acetate-buffered formulations, prediction of ²²⁷Th sorption was influenced by complex variable relationships and hence a risk of obtaining irreproducibility. Fine-tuned variable levels showed, however, variable combinations predicting high sorption of ²²³Ra (>90%) and low sorption of ²²⁷Th (<3%) also for the acetate-buffered formulations. The optimal separation conditions should be decided based on tuning the variables levels for ²²³Ra in the citrate-buffered formulations, while for acetate, the optimal separation should be based on tuning variable levels for ²²⁷Th sorption. The ionic strength of the formulation also seemed to affect the radionuclide sorption. Labeling of an antibody-chelator conjugate with purified ²²⁷Th (i.e. preparation of TTC) was successful in the selected citrate-buffered formulations tested.     

Systematic comprehensive techno-economic assessment of solar cooling technologies using location-specific climate data

A methodology for assessing solar cooling technologies is proposed. The method takes into account location specific boundary conditions such as the cooling demand time series, solar resource availability, climatic conditions, component cost and component performance characteristics. This methodology evaluates the techno-economic performance of the solar collector/chiller system. We demonstrate the method by systematic evaluation of 25 feasible combinations of solar energy collection and cooling technologies. The comparison includes solar thermal and solar electric cooling options and is extended to solar cooling through concentrated solar power plants. Solar cooling technologies are compared on an economic and overall system efficiency perspective. This analysis has implication for the importance of solar load fraction and storage size in the design of solar cooling systems. We also stress the importance of studying the relation between cooling demand and solar resource availability, it was found that overlooking this relation might lead to overestimations of the potential of a solar cooling system in the range of 22% to over 100% of the actual potential.