核燃料后处理厂钚尾端工艺方案的探讨

通过调研国内外后处理厂钚尾端工艺的生产流程,对批式和连续生产流程进行了对比,得出拟建的大厂钚尾端拟优先采用连续处理流程和设备。根据钚尾端连续流程特点,对设备的处理能力、运行稳定性、设备密封性和可维修等性能均有较高要求,对连续沉淀器、连续过滤机、连续焙烧炉、连续匀化器等主要设备从功能和工艺要求上进行了初步设想。     

Purex流程以N2O4调节钚价态的研究

详细描述了用液体N_2O_4调节Purex流程1BP、2BP、3BP和4W 料液中Pu(Ⅲ)价态的实验结果。研究酸度、钚离子浓度、铁离子浓度和温度等因素对N_2O_4氧化调价的影响。实验结果表明,N_2O_4氧化NH_2OH、N_2H_4和Pu(Ⅲ)-N_2H_4-NH_2OH的速度很快,在实验条件下是瞬间反应;在Pu(Ⅲ)-U(Ⅳ)-N_2H_4-HNO_3体系中,钚离子有明显的催化N_2O_4氧化U(Ⅳ)的作用;模拟Purex 流程1BP、2BP和3BP料液,用2.5倍化学计算量的N_2O_4,10—15℃下20min,可以达到Pu(Ⅲ)完全氧化到Pu(Ⅳ);调价后的料液放置5d,Pu(Ⅳ)的含量没有变化。     

国外核燃料后处理厂临界事故统计和分析

对国外核燃料后处理厂的临界事故进行了统计和分析。通过调研,明确了临界事故的实际发生频率、起因和危害,并分析了临界事故的预防、监测和缓解措施的投入及使用情况,以便为核燃料后处理厂相关的工程设计、设施运行、安全分析以及安全审评工作提供参考和依据。     

硝酸羟胺在动力堆乏燃料后处理流程钚线第三循环中的应用

为动力堆乏燃料后处理流程的设计之用，通过还原实验，研究温度、溶液酸度、硝酸羟胺（ＨＡＮ）与钚的初始浓度比值对ＨＡＮ还原Ｐｕ（Ⅳ）的还原百分数的影响。结果表明，升温、低酸和合适的ＨＡＮ用量有利于Ｐｕ（Ⅳ）的还原。用模拟料液进行３Ａ槽和３Ｂ槽串级实验的结果表明：３Ａ槽的钚回收率达９９．９％以上，３Ｂ槽的达９９．９８％；钚中去铀分离系数达５０左右。通过单级和串级实验，研究了含钚３０％（Ｖ／Ｖ）ＴＢＰ－煤     

Purex流程1AF料液中调节钚价态的研究 Ⅰ.二步法调价

在Ｐｕｒｅｘ流程１ＡＦ料液中调节钚的价态，先以ＮＨ＿２ＯＨ·ＨＮＯ＿３，Ｎ＿２Ｈ＿４·ＨＮＯ＿３和Ｕ（Ⅳ）为还原剂，详细研究了各种工艺变量对还原Ｐｕ（Ⅵ）的影响。其后，选择最佳还原条件，用模拟１ＡＦ料液做了验证。最终，完成了用ＮＯ￣－＿２氧化Ｐｕ（Ⅲ）至Ｐｕ（Ⅳ）的实验。     

核燃料后处理厂选址阶段的环境影响评价与分析

随着我国核电的高速发展,产生的乏燃料数量日益增多,建设工业规模的核燃料后处理厂是保证我国核能可持续发展的关键途径与步骤。由于我国尚未建立针对工业规模核燃料后处理设施的相关环保设施方案与环境影响评价标准体系,本文首次依据核燃料后处理厂可行性研究阶段的工程方案对运行状态和事故工况下的放射性流出物进行了估算,并参照我国核电环评相关标准规范对200吨/年核燃料后处理厂投运后可能对环境的影响程度进行了分析预测,可以作为审管当局决策的重要依据。     

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。     

乏燃料后处理厂主工艺中的主要化学安全问题

水法后处理工艺过程涉及很多化学反应,反应条件和反应产物不同,需要关注的化学安全问题也不同。描述了后处理主工艺不同阶段的化学反应,分析了各阶段应关注的主要化学安全问题,为商用后处理厂的设计和事故分析提供参考。     

乏燃料后处理萃取过程中的界面污物

在乏燃料后处理萃取工艺工程中，萃取剂和溶剂的辐解以及料液中不溶性固体微粒的存在导致产生界面污物。界面污物严重影响萃取柱的正常操作。文章就有关界面污物的研究状况进行概要评述。目前，普遍认为，在Purex流程萃取过程中，尤其是在一循环中，界面污物的产生与Zr和TBP降解产物HDBP、HzMBP、H3PO4形成的沉淀以及料液中存在的不溶性RuO2、Pd等微粒的表面化学现象有关。沉淀是否产生以及形成的界面污物类型与HDBP／Zr摩尔比和水相条件密切相关。此外，煤油等稀释剂的降解产物也是形成界面污物的一个不可忽略的因素。从萃取设备中排出的界面污物可用Na2CO3或草酸进行处理。在界面污物模拟实验中，需同时考虑Zr与TBP降解产物HDBP、H2MBP、H3PO4形成沉淀和不溶性微粒RuO2、Pd的影响等多种因素。     

Plutonium management with thorium-based fuels and inert matrix fuels in thermal reactor systems

The plutonium that is produced by light water reactors worldwide is currently re-used to a limited extent. In the last century, the expected introduction of fast reactors and the associated need for large amounts of plutonium did not take place. The result is that worldwide a stockpile of excess plutonium has formed, which is the dominant contributor to the radiotoxicity of spent nuclear fuel for storage times from 10² to 10⁵ years. One option to reduce or stabilize the plutonium stockpile is to utilize this plutonium in advanced fuel types, such as thorium-based and inert matrix fuels. Because these fuels do not contain uranium, the plutonium consumption rate is very high. In this paper, the status of the fuel research and some recent developments are given.     

普雷克斯流程中Np的行为和控制

在后处理流程的众多化学分离中 ,Np的走向和控制是国际后处理界关注的重点研究课题。根据我国和其他国家的研究成果 ,综合分析了后处理中Np的走向和控制。Np在辐照燃料溶解液中的价态分布主要取决于溶解液中HNO3与HNO2 之比 ,通常情况下 ,溶解液中Np(Ⅴ )占主要份额 ;Np在共去污阶段的走向有两种可能 ,一是将Np控制为Np(Ⅴ ) ,使其进入高放废液 (1AW ) ,二是将Np控制为Np(Ⅵ ) ,则Np将与U ,Pu一起进入有机相 ,但两者至今为止都难以实现定量分离。Np在U/Pu分离阶段部分随U ,部分随Pu。在U纯化循环中 ,理想的方法是采用低酸加热氧化Np(Ⅳ )至Np(Ⅴ ) ,以实现与铀的有效分离。     

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.     

Formation of Precipitate in High-Level Liquid Waste from Nuclear Fuel Reprocessing

Stability of high-level liquid waste (HLW) from nuclear fuel reprocessing was studied by using a simulated HLW. Fundamental works disclosed that precipitates formed during aging at ambient temperature or refluxing the simulated HLW in 2 mol/lHNO₃ solution consist mainly of Mo, Zr and Te contributing significantly to the formation of precipitate. When the simulated HLW was denitrated with formic acid or deacidified with NaOH, fractions of precipitated Mo, Zr and Te increased with pH and amounted to over 85% at pH 0.5, where the fraction of precipitated La was below 0.1%. For further treatment of HLW such as partitioning, denitration of HLW to pH 0.5 might be useful for removing Mo, Zr and Te from the solution without significant contamination with rare earths, Am and Cm.     

Pyrochemical Reprocessing of Weapons Plutonium as Nuclear Fuel for Power Reactors

Experimental work on transferring compact weapons plutonium by pyrochemical methods into its compounds was generalized. Powdered plutonium dioxide was obtained by interaction with moist air at 480–500°C. PuH₃ was synthesized by interaction of the metal with hydrogen at 280°C. PuF₃ was formed in an interaction of dioxides with HF and H₂. The reaction of plutonium hydride with HF also resulted in plutonium trifluoride. PuF₄ was synthesized by interaction of PuO₂ with HF.In summary, compact plutonium can be transfered by water-free methods into its powdered compounds, which can be used in investigations for developing new types of fuel elements for power reactors. 6 figures, 2 tables, 16 references.     

压水堆核电厂乏燃料后处理平准成本分析

基于压水堆核电厂乏燃料后处理工厂的规模与建设费用的半定量关系,以年处理能力为800 t的商用后处理厂为例,采用自上而下的方法,重点分析了隔夜成本、建造期和建造期间的利率,对采用PUREX流程的后处理工厂的建设费用影响;并用与建造期相同的利率,计算了UOX乏燃料后处理的平准成本.