氯化锂熔盐中八氧化三铀电解还原基础研究

将氧化物乏燃料直接电解还原为粗金属的过程是目前以电解还原-电解精炼为特征的主流干法后处理流程的重要步骤。二氧化铀（UO₂）是乏燃料的最主要成分,将致密的UO₂芯块转化为八氧化三铀（U₃O₈）粉末后,再进行电化学还原能有效提高还原速率。因此,以U₃O₈为研究对象,开展其在氯化锂（LiCl）熔盐中的电解还原机理研究,对后处理干法流程的开发具有重要的现实意义。本文在650 ℃的LiCl熔盐中,采用循环伏安法和恒电位电解法,研究U₃O₈的电解还原行为;对电解后的样品,运用XRD、SEM等手段分析其组成和形貌,并推测相应的还原机理。     

熔盐电解法乏燃料干法后处理技术研究进展

熔盐电解法是目前最有前途的干法后处理技术,适合于处理氧化物和金属等不同类型乏燃料。熔盐电解法主要包括四个核心流程,即首端处理、电解还原、电解精炼和提取以及废物处理。本文以国际上最新的研究进展为蓝本,综述熔盐电解法乏燃料后处理技术的基本流程以及待解决的关键问题。     

干法后处理流程中电解精炼设备研发进展北大核心CSCD

高燃耗快堆乏燃料具有高钚含量、强放射性、高释热率等特点。基于溶剂萃取原理的水法后处理工艺存在溶剂易辐解等问题,宜对高燃耗快堆乏燃料采用干法后处理工艺进行处理。熔盐电解干法工艺采用耐辐照的无机盐为介质,通过电化学方法分离回收锕系元素,是最具应用前景的干法后处理技术。在熔盐电解干法工艺流程中,承担锕系元素分离任务的电解精炼单元是核心环节。本文调研了乏燃料干法后处理过程中电解精炼设备的研发进展,分析了电解精炼设备关键技术和发展趋势,为我国快堆乏燃料电解精炼设备的研发提供了参考。     

俄罗斯氧化物乏燃料电沉积流程研究进展

高温熔盐干法后处理以熔盐作为电解质，通过电解精炼和电沉积回收核燃料中的铀和钚。目前，俄罗斯、美国、日本、韩国和欧盟等国均在积极发展乏燃料高温熔盐干法后处理技术的研究，其中俄罗斯的金属氧化物核燃料电沉积流程是经典的流程之一。本文对俄罗斯原子反应堆研究所（Research Institute of Atomic Reactors,RIAR）发展的氧化物乏燃料高温熔盐电沉积干法后处理的发展现状、流程及特点进行了综述。     

俄罗斯氧化物乏燃料电沉积流程研究进展

高温熔盐干法后处理以熔盐作为电解质,通过电解精炼和电沉积回收核燃料中的铀和钚。目前,俄罗斯、美国、日本、韩国和欧盟等国均在积极发展乏燃料高温熔盐干法后处理技术的研究,其中俄罗斯的金属氧化物核燃料电沉积流程是经典的流程之一。本文对俄罗斯原子反应堆研究所(Research Institute of Atomic Reactors, RIAR)发展的氧化物乏燃料高温熔盐电沉积干法后处理的发展现状、流程及特点进行了综述。     

铝合金化技术在乏燃料干法后处理中的应用研究进展

干法后处理在未来先进核燃料循环中将发挥关键作用。由美国开发的熔盐电精炼流程是目前最具应用前景的干法后处理流程之一,但是锕系元素(An)与镧系元素(Ln)的高效分离仍然是该流程目前亟待解决的关键科学与技术问题之一。研究表明,An与Ln形成铝合金时沉积电位差较大,采用固态铝电极电解有望实现An与Ln的有效分离,从而更好地服务于分离-嬗变策略。本文针对铝合金化技术在乏燃料干法后处理中的应用研究进展进行综合阐述,重点介绍铝合金化在熔盐电精炼中的应用研究,主要包括Ln和An的铝合金化行为、An和Ln的铝合金化分离等几个方面。     

干法后处理熔盐电解精炼过程数学模型研究

熔盐电解精炼是乏燃料干法后处理的核心工艺单元,通过数学模型探索高温熔盐电解精炼过程的化学与电化学变化,可为电解精炼工艺优化和设备设计提供参考依据。本文基于电化学热力学及物质传递公式建立了乏燃料熔盐电解精炼过程的数学模型,以铀钚锆三元合金燃料为研究对象,计算了燃料中关键元素的电极电势、分电流及物料分布随时间的变化。采用向后差分法对物料分布变化方程进行离散,通过文献实验数据对建立的数学模型进行了准确性验证。结果表明,模拟计算所得阴极沉积铀产品与实验数据的相对误差为2.80%,所建数学模型具有较好的拟合性。同时采用所建模型模拟计算了电流强度对乏燃料电解精炼过程的影响,结果表明电解速率与电流强度呈正比,不改变钚铀锆的溶解和沉积顺序。     

干法后处理技术典型流程综述

虽然基于溶剂萃取的Purex流程在乏燃料后处理几十年的应用中取得的成功,使得水法后处理至今没有发展出可以取代这一流程的新萃取剂,但干法后处理却有了两种可供进一步发展的流程:氟化物挥发法和高温电化学法。氟化物挥发法存在的最大问题是热力学上PuF6必须在有大量F2过剩的条件下才稳定。高温电化学法适合于处理合金元件,以及氧化物和碳化物元件。首先,将核燃料熔解在熔盐中,然后,电解使铀钚在阴极上沉积,再对阴极上沉积出来的铀钚进行精制而得到铀钚产品。但该方法存在熔盐对MOX的熔解能力和对过程设备的腐蚀问题。     

影响CaCl2-NaCl熔盐电脱氧法制备金属Zr的工艺条件

将氧化物转化为金属是熔盐电解精炼干法后处理氧化物乏燃料流程的关键步骤之一。在等摩尔CaCl₂-NaCl混合熔盐体系中,以石墨棒为阳极,采用高温烧结后的ZrO₂模拟UO₂开展了电脱氧制备金属Zr的FFC剑桥工艺条件优化。研究了工艺条件（槽电压、电解时间、烧结温度和电解温度等）对电脱氧制备Zr的影响。采用场发射扫描电子显微镜(SEM)和X射线衍射(XRD)分别分析了电解前后ZrO₂阴极的微观结构和物相组成。优化后的工艺条件为：电压3.4 V、电解时间12 h、烧结温度900 ℃和电解温度722 ℃。同时,研究结果表明, ZrO₂电脱氧还原为Zr时,存在中间产物CaZrO₃和ZrO。     

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。     

高温气冷堆乏燃料后处理首端技术研究进展

高温气冷堆乏燃料采用后处理路线能充分利用核资源并减少需要最终地质处置的核废物量,有利于核能的可持续发展。传统的LWR乏燃料后处理首端过程不能用于处理高温气冷堆的乏燃料。高温气冷堆乏燃料元件及包覆层颗粒的破碎是首端处理技术的难点。破碎乏燃料元件及去除石墨的方法主要有机械碾碎法、燃烧法、脉冲电流法等;破碎及去除碳化硅的方法有传统机械碾碎法,以及正在发展中的熔融法、气流喷射粉碎法等,其中,气流喷射粉碎法具有较好的发展前景。目前,尚无一种理想的技术来解决高温气冷堆乏燃料后处理中的首端过程问题,需进一步开展高温气冷堆乏燃料后处理技术的研究。     

Actinide recycling by pyro-process with metal fuel FBR for future nuclear fuel cycle system

Pyro-metallurgical technology is one of potential devices for future nuclear fuel cycle. Not only economic advantage but also environmental safety and strong resistance for proliferation are required for the fuel cycle. In order to satisfy the requirement, actinides recycling applicable to LWR and FBR cycles by pyro-process has been developed since more than ten years in CRIEPI. The main technology is electrorefining for U and Pu separation and reductive-extraction for TRU separation, which can be applied on oxide fuels through reduction process as well as metal fuels. The application of this technology on separation of TRU in HLLW through chlorination could contribute to the improvement of public acceptance on the geologic disposal.

The main achievements are summarized as follows:

• -|The elemental technologies, such as electrorefining, reductive extraction, injection casting and salt waste treatment and solidification, have been developed successfully with lots of experiments

• -|The fuel dissolution into molten salt and uranium recovery on solid cathode for electrorefining have been demonstrated by engineering scale facility in Argonne National Laboratory by using spent fuels and in CRIEPI by uranium tests.

• -|Single element tests, using actinides, showed the Li reduction to be technically feasible, remaining the subjects of technical feasibility on multi-elements system and on effective recycle of Li by electrolysis of Li₂O.

• -|Concerning on the treatment of HLLW for actinide separation, the conversion to chlorides through oxides has been also established through uranium tests.

• -|It is confirmed that more than 99% of TRU nuclides can be recovered from the high level liquid waste by TRU tests

• -|Through these studies, the process flow sheets for reprocessing of metal and oxide fuels and for partitioning of TRU separation have been established.

The subjects to be emphasized for further development are classified into three categories, that is, process development (demonstration), technology for engineering development, and supplemental technology.

The metal fuel FBR has a high potential for recycling actinides by integration with pyro-reprocessing. Alloys of U-Pu-Zr with minor actinides are investigated from points of fuel properties. The miscibility and other characteristics suggest that the maximum content up to ca. 5 wt% of minor actinides is allowable in the matrix. Nine pins of metal fuel including minor actinides are ready for irradiation at Phenix fast reactor.     

Electrochemical Properties of Uranyl Ion in Ionic Liquids as Media for Pyrochemical Reprocessing

We have proposed a new reprocessing process by using ionic liquids (ILs) instead of molten salts of alkali chlorides in pyrochemical process. In the proposed process, spent nuclear fuels are dissolved in ILs by using Cl₂ as an oxidant, and UO₂ ²⁺ and PuO₂ ²⁺ ions in ILs are recovered as UO₂ and PuO₂ by electrochemical reduction. In order to examine applicability of ILs as media for reprocessing, we have studied electrochemical behavior of UO₂ ²⁺ in 1-butyl-3-methylimidazolium chloride (BMICl), 1-butyl-3-methylimidazolium tetrafluoroborate (BMIBF₄), and 1-butyl-3-methylimidazolium nonafluorobutanesulfonate (BMINfO). Electrochemical properties of uranyl chloride dissolved into ILs were examined by cyclic voltammetry. In BMICl, an almost reversible redox couple was observed, and the formal potential and the diffusion coefficient were evaluated as _0:758V vs. Ag/AgCl and 4:8 × 10^?8 cm²s^?1, respectively. On the other hand, the electrochemical reactions of UO₂ ²⁺ in BMIBF₄ and BMINfO were irreversible. In BMINfO, some reduction peaks and one sharp oxidation peak were observed in the range of ?0:6～–0:2V and around 0.85V vs. Ag/AgCl, respectively. The reduction and oxidation peaks were assigned to multi step reduction of UO₂ ²⁺ to U(IV) via U(V) and/or direct reduction of UO₂ ²⁺ to U(IV), and the oxidative dissolution of the resulting U(IV) compounds, respectively. The electrochemical reduction of UO₂ ²⁺ in BMINfO at ?1:0V vs. Ag/AgCl produced the deposits on a carbon electrode as a cathode. Analyses of the deposits with the scanning electron microscope and the energy dispersive X-ray spectrometer indicated that the deposits are compounds containing uranium, oxygen, and chlorine. As a result, it is expected that the UO₂ ²⁺ in IL can be recovered electrolytically as uranium compounds such as UO₂ and uranium oxychlorides.     

Electrolytic Reduction of Spent Light Water Reactor Fuel Bench-Scale Experiment Results

A series of experiments were performed to demonstrate the electrolytic reduction of spent light water reactor fuel at bench-scale in a hot cell at the Idaho National Laboratory Materials and Fuels Complex. The process involves the conversion of oxide fuel to metal by electrolytic means, which would then enable subsequent separation and recovery of actinides via existing electrometallurgical technologies, i.e., electrorefining. Four electrolytic reduction runs were performed at bench scale using ～500 ml of molten LiCl–1 wt% Li₂O electrolyte at 650°C. In each run, ～50 g of crushed spent oxide fuel was loaded into a permeable stainless steel basket and immersed into the electrolyte as the cathode. A spiral wound platinumwire was immersed into the electrolyte as the anode. When a controlled electric current was conducted through the anode and cathode, the oxide fuel was reduced to metal in the basket and oxygen gas was evolved at the anode. Salt samples were extracted before and after each electrolytic reduction run and analyzed for fuel and fission product constituents. The fuel baskets following each run were sectioned and the fuel was sampled, revealing an extent of uranium oxide reduction in excess of 98%.     

U(Ⅵ)的还原固定研究进展

铀是重要的天然放射性元素，也是最重要的核燃料。在铀矿选冶、核能发电及乏燃料后处理等过程中会产生一定量的含铀废水，对生态环境和人类健康造成潜在威胁。将易溶的U(Ⅵ)还原为难溶的U(Ⅳ)是处理含铀废水的常用方法之一。本文综述了三种常见的U(Ⅵ)还原方法，即零价铁还原、微生物还原及光催化还原，比较了三种方法的优缺点并对其未来应用前景进行了展望。     

A new concept of nuclear fuel reprocessing by applying ion-exchange technology

The development of advanced technology for the spent nuclear fuel reprocessing should be achieved not only considering cost, non proliferation and reduction of radioactive wastes but also corresponding to both spent nuclear fuels of LWR and FBR.

We have proposed an ion exchange process for reprocessing using a new type ion exchanger developed to chemical method of U enrichment technology. This process possess possibility of a sharp cut in cost, since this ion exchanger is characterized by rapid adsorption-desorption rate dominating the treatment rate.

From the basic experimental results, this reprocessing process has been constructed by 3 ion exchanger columns which consist of a main separation column, the uranium-refining column and the plutonium-refining column.

Comparing ion exchange process with the conventional Purex process, this ion exchange process has many advantages such as the decrease in the number and size separation equipment, solvent-spent free and alkaline-liquid-spent free. With these advantages, this process is estimated that the construction cost of reprocessing process is greatly reduced comparing to the conventional process.