离子液体浮选分离模拟乏燃料中的稀土元素

从乏燃料中高效分离稀土元素（中子毒物）是实现乏燃料再生循环利用的关键步骤。利用双有机相离子液体选择性浮选分离乏燃料中的稀土元素,使氧化铀和稀土氧化物几乎不被溶解,实现两者固相之间的分离,避免了二次废液的产生,具有节能和环保的双重意义。以2-乙基己基膦酸单-2-乙基己基酯（P507）为稀土元素的捕收剂、煤油或油酸为稀释剂,以1-丁基-3-甲基咪唑六氟磷酸盐（［C₄mim］［PF₆］）和1-丁基-3-甲基咪唑双三氟甲烷磺酰亚胺盐（［C₄mim］［Tf₂N］）两种离子液体作为浮选体系的另一相,优化得到了浮选分离稀土氧化物的最佳条件。结果发现：浮选分离去除率随着混合物中初始Nd含量以及浮选次数的增加而增加。另外,对所有稀土元素与U₃O₈分别组成的二元体系混合物进行了浮选分离研究,发现在相同条件下,该体系对不同稀土元素的分离也不同,浮选分离的去除率与稀土氧化物的密度有一定的相关性。在此基础上,利用浮选机开展了工艺化的初步探索,发现该浮选体系对Nd的去除率可达80%以上。     

离子液体体系的萃取行为及其在乏燃料后处理中的应用前景

离子液体由于其特有的性质,在乏燃料后处理中的应用已受到广泛关注。本文综述了不同种类离子液体中多种萃取剂对乏燃料所含若干锕系元素及放射性裂片元素的萃取,重点分析了不同萃取体系的萃取效率、萃取选择性、萃取机理和反萃等关键问题。综合目前的研究成果,可发现:离子液体-萃取剂体系由于其独特的萃取机理,通常比传统萃取体系具有更高的萃取效率;一些萃取体系具有高选择性使其在乏燃料后处理中有很好的应用前景。在简要介绍阳离子交换机理、阴离子交换机理和中性复合物机理三种离子液体体系萃取机理的同时,重点总结了萃取中三相问题和协同萃取效应。本文还总结了液-液反萃、超临界CO2反萃和电化学反萃三种常见的反萃方法,讨论了各自的优缺点。本文最后对离子液体在乏燃料中的应用进行了总结与前景展望。     

离子液体在金属离子萃取分离中的应用

离子液体作为绿色溶剂是溶剂萃取分离金属离子方面研究的热点.介绍了以离子液体为溶剂时,萃取分离各种金属离子的效果,包括碱金属、碱土金属、过渡金属、稀土及锕系金属,以及核燃料后处理涉及的铀、钚及裂变产物等多种离子,深入探讨了其萃取机理.展望了离子液体取代有毒、易挥发、易造成环境污染的有机溶剂的发展和应用前景.     

镧系及锕系元素在离子液体中的电化学行为

乏燃料回收是核燃料循环的核心,对核安全和核能可持续发展具有重要的意义,其分为使用水溶液的湿法和不使用水溶液的干法处理。熔盐电解技术是乏燃料干法回收的重要方法之一,但其工艺温度往往在数百摄氏度,对设备和能耗要求都很高。离子液体具有电化学窗口宽、低熔点、低蒸汽压、热稳定性好等优点,有望替代高温熔盐用于乏燃料干法回收。本文概述了镧系元素和锕系元素在离子液体中电化学方面的研究状况,表明离子液体用于乏燃料干法回收是可行的,但需要更多的基础性研究。     

功能离子液体在镧锕系离子萃取分离中的研究进展北大核心CSCD

离子液体因其独特的物化性质已成为溶剂萃取领域中绿色环保、极具应用前景的稀释剂。除了对萃取剂和萃合物有良好溶解性之外,离子液体萃取体系通过离子交换机理的协同作用,往往还展现出更优的萃取性能。而功能离子液体通常是在离子液体结构基础上进行化学修饰,使其兼具离子液体和螯合基团的双重功能,在萃取过程中不仅能用作稀释剂,还具有萃取作用。近十余年来,功能离子液体在放化分离领域已引起广泛关注。本文首先对离子液体和功能离子液体进行了简介,重点讨论其结构上的关联与区别;再以功能为导向,综述了功能离子液体在铀酰离子、钍离子、镧系和次锕系金属离子萃取分离方面的国内外研究现状,并对功能离子液体在镧锕萃取分离领域的未来研究趋势进行了展望。     

氧化铀在离子液体中溶解及铀的分离

传统干法后处理中常使用高温熔融盐溶解氧化铀,并通过电化学方法分离纯化。离子液体作为新型溶剂具有更低的熔点、更宽的电化学窗口和较好的溶解性,并且可进行设计修饰,在溶解氧化铀方面具有很好的应用前景。本文对离子液体体系溶解铀氧化物和铀的分离纯化进行了总结,并讨论了溶解体系的表征手段。     

乏燃料的干法分离技术

【日本《能源》2000年3月刊第48～50页报道】 分离转化为短半衰期核素 核电站(轻水堆)使用的核燃料(乏燃料)以被称为普雷克斯法的工艺进行后处理。通过这种方法就可从乏燃料中回收铀(U)、钚(Pu),这些回收物质又可再次在燃钚轻水堆和快堆中作为燃料利用。另一方面,把产生于后处理工序的高放废液做成玻璃固化体,然后对其进行深层处置。 高放废液中含有镎(Np)、镅(Am)、锔(Cm)等半衰期非常长的超铀元素,它们和半衰期短的其他核素一并被处置。现正在对高放玻璃固化体处置的安全性进行评估,这将成为今年处置工作的主体。如果能够除去来自高放废…     

离子液体在镧系锕系元素萃取分离应用中的理论计算研究进展北大核心CSCD

离子液体在核燃料后处理萃取分离过程中被认为具有一定的应用前景。常规的萃取剂在离子液体中展现出比传统溶剂体系更优异的萃取分离性能,因此离子液体自身在其中的角色和作用值得研究。本论文综述了以分子动力学模拟和量子化学计算方法研究离子液体体系萃取分离镧系锕系元素的相关研究工作,重点关注金属离子和金属配合物在离子液体中与离子液体阴阳离子之间的相互作用。     

氮杂多环芳烃/离子液体萃取体系的伽马辐射效应

乏燃料后处理分离体系的辐射稳定性是其实际应用前需解决的重要问题。采用傅里叶变换红外光谱(FTIR)、超高效液相色谱/四级杆飞行时间质谱联用(UPLC/Q-TOF-MS)等方法系统研究了IB-BTP/[C₂mim][NTf₂]、CA-BTP/[C₂mim][NTf₂]和CA-BTPhen/[C₂mim][NTf₂]三种典型的氮杂多环芳烃/离子液体萃取分离体系的γ辐射效应,并通过Eu³⁺萃取实验对辐照前后体系的萃取性能变化进行了对比。结果表明:三种萃取剂在[C₂mim][NTf₂]离子液体中的辐射稳定性顺序为:CA-BTP>IB-BTP≈CA-BTPhen;三种体系的辐解产物主要为[C₂mim][NTf₂]离子液体辐解产生的·CF₃、·[C₂mim]⁺、·H等自由基进攻氮杂多环芳烃...     

模拟乏燃料的氧化挥发首端工艺研究

通过实验室制备的UO₂模拟芯块,分别研究了氧化与还原气氛下,温度、气体组成和保温时间对粉化与转化过程的影响,结果显示,氧化条件为空气/450℃/4 h、还原条件为4%(体积分数)H₂-Ar/700℃/4 h的三次氧化还原循环流程,对UO₂模拟芯块和真实天然铀芯块均有良好的粉化效果。针对制成的包含有多种裂变元素的模拟乏燃料,在经过三次氧化还原循环流程处理的基础上,进一步结合1 200℃/4 h的更高温挥发技术,形成国内首个模拟后处理氧化挥发首端工艺。该工艺能够使Mo、Te、Se和Ru等半挥发性裂变元素以氧化物的形态被有效去除,去除率均达到85%以上。     

离子液体/苯并15-冠-5浸渍XAD-7树脂萃取分离锂同位素

以XAD-7树脂为支撑担体制备了含有三种不同咪唑型离子液体([C_8mim][BF_4]、[C_8mim][PF_6]、[C_8mim][(SO_2CF_3)_2N])和萃取剂(苯并15-冠-5)的浸渍树脂,并用于锂同位素的萃取分离。浸渍树脂的红外和扫描电镜表征表明,离子液体成功负载到了树脂上;热重分析表明,该浸渍树脂具有良好的热稳定性。在水相初始pH=5.55时,浸渍树脂具有最佳萃取率。浸渍树脂在LiSCN溶液中具有较高的萃取率,而在CF_3COOLi溶液中呈现较大的单级分离因子,最大单级分离因子达到1.045±0.002。浸渍树脂的萃取平衡时间为2.5~3h。萃取热力学研究表明,该反应为自发过程,温度对体系的影响较小。~6Li富集于固相,~7Li富集在水相。该系列浸渍树脂易于再生,可循环使用。     

The Application of an Advanced Ion Exchange Process to Reprocessing Spent Nuclear Fuels, (I)

In order to develop an advanced ion exchange process for the reprocessing of spent nuclear fuels, a novel anion exchanger, AR-01 with the resin embedded in porous silica beads and benzimidazoles as functional groups has been manufactured. Adsorption behavior of various fission product elements (FPs) and uranium in nitric acid medium were investigated experimentally using this anion exchanger. Separation performance of FPs from U(VI) in simulated spent fuel solutions was demonstrated by column chromatography utilizing dilute HNO₃ and thiourea as eluents.

Most FPs such as Cs(I), Sr(II), Mo(VI), Rh(III) and trivalent rare earths showed negligibly slight adsorption and could be separated from U(VI) satisfactorily. Cerium(IV) was strongly adsorbed, but was gradually reduced to non-adsorptive Ce(III) by the anion exchanger. Zirconium(IV) presented weak adsorption and its a part mixed with U(VI) in the column experiments. Ruthenium(III) exhibited quite strong adsorption in a broad HNO₃ concentration range as the form of anionic nitrosylnitrato-complexes, its most amount mixed with U(VI). Palladium(II) showed significantly strong adsorption probably due to complexes formation with the anion exchanger. The adsorbed Pd(II) was effectively eluted out by thiourea and separated from U(VI) and other FPs completely.     

增压排空阴离子交换法分离-ICP-MS测定铀基体中痕量稀土

采用增压排空的阴离子交换柱实现了铀基体与痕量稀土元素的分离。结果表明,全流程对稀土的回收率在96%～106%之间,整个分离过程可在20min内完成。用ICP MS测量铀基体中痕量稀土的检测限为1×10-12g/mL。用该方法分析U3O8成分分析标准参考物质GBW04207中的痕量稀土含量,测量结果与参考值符合良好。     

Recovery of Transuranium Elements from Real High-Level Liquid Waste by Pyropartitioning Process

A pyropartitioning process is under development to recover minor actinide elements from high-level liquid waste (HLLW) generated by Purex reprocessing. This pyropartitioning process consists of a denitration step that converts various elements in the HLLW into oxides, a chlorination step that converts the oxides into chlorides, and a reductive-extraction step that separates the actinide elements from fission products (FPs) in the chlorination product. The feasibility of each step was confirmed using simulating FPs and unirradiated transuranium elements (TRUs). In the present study, approximately 520 g of real HLLW was prepared to demonstrate the feasibility of the pyropartitioning process. Almost 100% of each TRU originally contained in the HLLW was recovered in the liquid cadmium phase in the reductive-extraction step, which showed that the expected chemical reactions were completed and the mass loss of TRUs was negligible in the denitration, chlorination, and reductive-extraction steps. The separation behaviors of actinide elements, including americium and curium, from FPs in the reductive-extraction step were quite similar to those observed in previous experiments using unirradiated materials. Hence, the pyropartitioning process was successfully verified.     

Thermochemical Modeling of Electrorefining Process for Reprocessing Spent Metallic Fuel

Numerical simulation of the electrotransport of fuel actinides, minor actinides and rare earths in an electrorefiner for pyrochemical reprocessing of a typical spent metallic fuel has been attempted based on an improved thermochemical model developed for application to a multi-component system in the electrorefiner. Variation of the decontamination factors (DFs) for minor actinides and rare earths and minor actinides fraction in the cathode product (MA,) as a function of some important parameters such as percentage transport, catholyte concentration, fuel composition (with respect to U and Pu) and batch number (batch processing) are presented and discussed.     

Equilibrium Distributions of Actinides and Fission Products in Pyrochemical Separation Systems, (I)

A thermodynamic model has been described for predicting the equilibrium distributions of solute elements in pyrochemical separation systems of molten salts and liquid metals. The model assumes that the solutes are present in the form of complex compounds in the salt phase and of intermetallic compounds in the metal phase. The thermodynamic data of these compounds are available in the literatures and also can be estimated theoretically. Some predictions have been made by the present model to discuss the feasibility of liquid metal extraction system for the recovery of platinum group elements from spent fuel.     

Electrodeposition of Uranium and Transuranic Elements onto Solid Cathode in LiCl-KCl/Cd System for Pyrometallurgical Partitioning

A pyrometallurgical partitioning process is being developed for recovering transuranic elements (TRUs) from high-level liquid waste. In the process, actinides are separated from fission product, especially rare earth elements (REs), by means of an electrorefining technique or a reductive-extraction technique. In this study, electrorefining experiments were carried out in LiClKCl/Cd system to recover actinides from salt bath containing actinides and REs. Uranium and neptunium could be depleted from the salt bath and recovered onto a solid cathode with high collection efficiency and high selectivity. Plutonium and americium, however, were difficult to be recovered at high current efficiency because reduction of Nd³⁺ to Nd²⁺ at about—1.7V consumed cathodic current prior to the deposition of Pu or Am. The rotation of the cathode had rather negative effect against deposition of Am and Pu in case of coexistence of much amount of Nd because Nd²⁺ was removed from the cathode surface quickly and the reaction of Nd³⁺ to Nd²⁺ was promoted. At higher current density, Pu and Am could be recovered onto solid cathode but current efficiency became too low. The result indicated that electrorefining technique in the pyro-partitioning was effective for U and Np but not for Pu and Am.