Recovery of Uranium and Plutonium from Spent Fuel Elements of Nuclear Reactors

The results of investigations of the preliminary removal of the products of radioactive decomposition from irradiated nuclear fuel to obtain uranium and plutonium which are suitable for reuse in fuel fabrication are presented. Nitrate-alkali melts are used for the operation. The experiments are performed on simulators and irradiated samples of BOR-60 fuel in remote-controlled hot boxes. The coefficients of removal of fission products are presented. A technological scheme, which will shorten the fuel cycle, for purifying hot nuclear fuel is recommended. __________ Translated from Atomnaya Energiya, Vol. 99, No. 5, pp. 387–392, November, 2005.     

核燃料后处理厂钚的水解聚合及应对措施

核燃料后处理是提取钚的重要途径,钚在后处理PUREX工艺流程中主要以溶液形式存在。钚溶液化学行为非常复杂,工艺运行过程中疏忽或瞬间不稳定的条件都有可能导致钚的水解和聚合。而聚合物一旦形成就很难破坏,会严重干扰萃取分离等工艺指标,同时也会导致潜在的工业安全问题和核临界安全风险。本文介绍了钚水解聚合领域的研究结果,结合核燃料后处理工艺的特点,分析了钚水解聚合的影响因素和安全风险,并提出了应对措施。     

Properties of Plutonium Dioxide as Nuclear Fuel

Generalized structural, thermodynamic, thermophysical, and strength characteristics as well as data on the dimensional stability, compatibility, and fragment gas release under irradiation of plutonium dioxide, used as a compact nuclear fuel for fuel elements in fast research reactors, are presented. Reliable operation of reactors with fuel elements based on plutonium dioxide confirmed that the properties of the fuel ceramic of fuel elements with austenitic stainless steel cladding have been determined with a high degree of reliability. 7 figures, 6 tables, 23 references.     

Crud in Solvent Washing Process for Nuclear Fuel Reprocessing

In a solvent washing process for nuclear fuel reprocessing, one of the important problems is a formation of stable emulsions between organic and aqueous phases. These emulsions are called interfacial “crud”. Crud is defined as an emulsion stabilized by finely dispersed solids. These stable emulsions lead to decreased washing efficiency, lower phase separation, disturbance of the interfacial control at the settler of the extractor, and so on. Cruds formed by precipitates of Zr and tributyl phosphate (TBP) degradation products, such as di-n-butyl phosphate (HDBP), mono-n-butyl phosphate (H₂MBP), and phosphoric acid (H₃P0₄) are studied by experiments using a sodium carbonate solution as a washing reagent. Experimental results show that not only pH value of the washing reagent, but also phosphate and zirconium mole ratio (P/Zr) are important in crud formation. Moreover, it is shown that the complex of Zr and HDBP, or Zr and H₂MBP has a significant role in stabilizing emulsions. However, the complex of Zr and H₃P04 is not effective in stabilizing cruds.     

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.     

乏燃料后处理厂草酸钚沉淀器临界安全分析研究

本文对乏燃料后处理厂中钚尾端工艺环节的关键设备草酸钚沉淀器进行了临界控制方法和参数的详细分析。针对连续沉淀器的工艺和结构特点,对易裂变物质的状态进行了一系列分析,比较了均匀溶液和悬浮颗粒溶液反应性的差别。对单个沉淀器和多个沉淀器并行工作的情况分别进行了临界安全分析,并分别研究了不含中子毒物、布置中子毒物层以及布置中子毒物棒等情况下能达到的最大处理能力。选取了临界安全基准实验国际评价中的相似实验方案进行了验证计算,分析了所用程序计算此类问题的不确定度。本文开展的临界安全分析研究总结了连续沉淀器临界安全控制的规律性结论,可为后续连续沉淀器的工艺设计及今后的工程应用提供参考。     

Low-Temperature Partitioning of Plutonium with High U(VI)-Loading for Fuel Reprocessing

Measurement of the distribution ratios of Pu(IV), U(VI) and HNO₃ at low temperatures and its treatment with DIST code revealed that a high U (VI)-loading of 30% TBP in n-dodecane splits Pu(IV) down to the aqueous phase more strongly than do at 25°C. Based on these findings, flowsheet conditions to separate Pu(IV) from U(VI) were investigated with EXTRA.M code including the distribution equations obtained above. And tentative flowsheets for non-reductive Pu-splitting process at a temperature of 5°C were proposed for fuel reprocessing mainly based on the effects of U (VI)-loading in the solvent and temperature on distribution ratios of Pu(IV) and U(VI). Distribution ratios of the fission products, Zr, Nb, Ru and Ce were also measured to assess their decontamination from U or Pu products in the above process. Finally behavior of Np, in the proposed partitioning process was discussed by analysis with EXTRA. M code and a redox reaction model.     

乏燃料干法后处理研究中的分析方法

综述了几种典型的乏燃料干法后处理方法,并对其中使用的分析方法进行了总结。详细论述了干法后处理研究中的在线分析方法,包括电化学分析方法、紫外可见吸收光谱法、X射线衍射法、拉曼原位分析、EXAFS原位分析、NMR原位分析等。在线分析方法有助于对工艺料液中物质的形态及结构进行实时监测。此外,离线分析方法可作为在线方法的有效补充,根据研究对象的形态(气态、液态、固态)对一些典型的离线分析方法进行了论述。     

核燃料循环中萃取剂的辐射化学研究进展

本文主要综述了30年来我国科研工作者在核燃料循环方面对萃取剂的辐射化学研究中所取得的一些研究成果,其中主要介绍了磷酸三正丁酯(TBP)萃取剂的辐射化学研究.此外还介绍了其它新型萃取剂的辐射化学的国内外研究现状.最后对该研究领域所面临的挑战和前景进行了讨论.     

Major Breakthrough in China~'s Nuclear Fuel Reprocessing Technology

正On November 25,the advanced process for special assembly reprocessing and hot test program,which was conducted by the CIAE~'s Department of Radiochemistry,passed the sci-tech achievement appraisement organized by the State Administration     

Spent Nuclear Fuel Reprocessing and Nuclear Materials Recycling in Two-Component Nuclear Energy

Atomic Energy - The project Proryv [Breakthrough] now being implemented in our country is aimed at achieving a new quality of large-scale nuclear power, development, origination, and industrial...     

乏燃料干法后处理技术研究进展

本文介绍了近年来各国的干法后处理研究计划,对干法后处理技术路线、流程特点和发展现状进行了综述.     

乏燃料后处理厂核材料衡算仿真

为优化乏燃料后处理设施的核材料衡算,寻找核材料衡算不平衡差（MUF）的主要因素,采用基于数值模拟的系统仿真方法,以核材料衡算视角构建乏燃料后处理设施核材料衡算仿真模型。改变模型工艺参数仿真不同规模的后处理设施中各环节核材料的流通量,然后以正态分布随机变量模拟各铀钚衡算测量点的随机误差,将这些带有随机特征的测量值叠加相应测量的系统误差作为核材料的仿真测量值。仿真计算结果表明,1AF中Pu、U含量测量的系统误差的方差分别占整体MUF方差的50%、40%以上,是主要误差来源。1AF的体积测量误差较小,占比MUF方差小于15%。废液中U和Pu含量很低,U和Pu含量测量的误差分别为10%和30%,对MUF方差影响不大,占比MUF方差分别小于3%和1%,废液的体积测量误差较小,占比MUF方差小于1%。U和Pu产品测量误差的方差占比MUF方差界于1AF和废液的测量之间,不是MUF误差的主要来源。     

后处理工艺料液中亚硝酸的荧光分析方法

建立了硝酸介质中测定亚硝酸的荧光分析方法。该方法以5-氨基荧光素作荧光增强剂,在酸性条件下,5-氨基荧光素与亚硝酸发生重氮化反应,其产物在碱性条件下具有很强的荧光,利用荧光光度计测定其荧光强度从而得到亚硝酸的含量。实验表明,样品测量的精密度为7%,回收率为89%～104%,样品的检测限为7×10-5mol/L,测量的不确定度为13%。该方法灵敏度高、操作简便、抗干扰强,以硝酸为介质,可直接用于后处理工艺中亚硝酸的测定。