高放废液的处理处置是影响核能可持续发展的重要因素之一.从高放废液中萃取分离高释热放射性核素137 Cs不仅有利于实现高放废液的安全处置,也可满足137 Cs在诸多工业领域的应用需求.研究提出了以杯芳烃冠醚衍生物为萃取剂从高放废液中萃取分离Cs的工艺流程,并分别采用模拟和真实高放废液对流程进行了验证实验.结果表明,模拟高放废液实验中Cs(Ⅰ)的萃取率达到99.9％,热实验中137 Cs的萃取率达到99.95％.该工作所提出的工艺流程为进一步开展我国动力堆高放废液处理及137 Cs分离提取提供了参考数据.

We propose a method to extract Sr-90 from high-level liquid waste (HLLW) using N,N,N',N'-tetraoctyl as the extractant and tri-n-butyl phosphate as the phase modifier. The process was tested with real HLLW with a recovery percentage of 99.91% of Sr-90. The results show that the proposed process for Sr extraction is efficient in the treatment of commercial HLLW and separation of Sr-90.

99Tc, a long-lived(T1/2=2.13×105 a) nuclide, has a complex routing in PUREX process due to its various oxidation in the reprocessing of the spent nuclear fuel which gets special attention. Rhenium is usually used as a simulant to study the chemical behavior of technetium since they have similar chemical behaviors. The electrochemical behavior of ReO-4 in 0.5 mol/L HClO4 solution was studied by cyclic voltammetry, differential pulse voltammetry and chronopotentiometry. The result shows that there is a multi-step electron transfer process in the electrochemical reaction of rhenium. Re(Ⅶ) is easily reduced to Re(Ⅳ) by three-electron transfer on platinum electrode, on which Re(Ⅶ) can also be gradually reduced to Re(Ⅵ) and Re(Ⅴ) through single electron transfer. When the reduction potential is lower than 0 V, Pt-Had is electrochemically formed on the platinum electrode surface and subsequently reduce Re(Ⅳ) to Re(Ⅲ). In addition, the electrochemical study of rhenium on gold electrode shows that Re(Ⅶ) is gradually reduced to Re(Ⅲ) by multi-step electron transfer and the Re(Ⅲ) can be further deposited as rhenium.

99Tc is one of the main fission products in nuclear spent fuel in power reactor. Its half-life period is 2.13×105 years, which has long-term harm to the ecological environment, and it gets special attention in PUREX process. The chemical properties of rhenium are similar to technetium. Studying the electrochemical behavior of rhenium is of great reference value for understanding the chemical properties of technetium. The electrochemical behavior of perrhenate in sulfuric acid and nitric acid on platinum electrode is studied. The results of cyclic voltammogram and bulk electrolysis with coulometry show that the reduction of ReO-4 ions on the platinum electrode is a multistep electron transfer process, and ReO2 is deposited on Pt electrode. Then the redox reaction of ReO-4/ReO2 will be established. In nitric acid, when the concentration of nitric acid is 1.00 mol/L, the reduction process of ReO-4 is similar to that in sulfuric acid system. When the concentration of nitric acid is 2.00, 4.00 mol/L respectively, nitric is reduced to nitrite by electrochemical reduction, which oxidizes the low valence state rhenium in solution.

研究了磷酸三丁酯(TBP)辐解产物磷酸二丁酯(HDBP)和磷酸一丁酯(H2 MBP)对U(Ⅳ)-肼以及乙异羟肟酸(AHA)反萃Pu(Ⅳ)的影响,考察了相接触时间、相比(o:a)、还原剂浓度、HNO3浓度、肼浓度、TBP辐解产物HDBP和H2 MBP浓度等条件对含有HDBP或H2 MBP的30％(体积分数)TBP/煤油中Pu(Ⅳ)反萃率的影响.结果表明:U(Ⅳ)对Pu(Ⅳ)有很强的还原反萃能力,降低相比、HNO3浓度、肼浓度有利于U(Ⅳ)对Pu(Ⅳ)的反萃,并且U(Ⅳ)可以快速有效地破坏HDBP、H2MBP与Pu(Ⅳ)的络合,将Pu(Ⅳ)反萃到水相.乙异羟肟酸对Pu(Ⅳ)有很强的络合反萃能力,通过降低酸度、延长相接触时间和增大AHA浓度能够有效降低HDBP和H2 MBP对AHA络合反萃Pu(Ⅳ)的影响.

N,N,N'N'-四辛基-3-氧戊二酰胺(TODGA)是最常用的酰胺荚醚类萃取剂,N,N,N'N'-四辛基-2-甲基-3-氧戊二酰胺(Me-TODGA)是TODGA的一种衍生物,有望取代TODGA应用于高放废液分离流程.本文对比研究了Me-TODGA和TODGA萃取三价锕系、镧系元素以及Sr(Ⅱ)、Zr(Ⅳ)、Pd(Ⅱ)等裂片元素的性能,并考察了吸收剂量对两者萃取性能的影响.结果表明,Me-TODGA和TODGA萃取高浓度金属离子时会出现三相,加入0.50 mol/L TBP-煤油溶液作相改良剂时,可防止三相的出现;Me-TODGA/TBP体系对于锕系、镧系、裂片元素的萃取能力弱于TODGA/TBP体系,有利于三价锕系和镧系元素反萃以及裂片元素的洗涤.Me-TODGA和TODGA萃取Eu(Ⅲ)和Sr(Ⅱ)分配比的对数值都随吸收剂量线性下降,但Eu(Ⅲ)和Sr(Ⅱ)的分离因子随吸收剂量变化较小.Me-TODGA的辐照稳定性相对于TODGA较差,加入TBP后体系的耐辐照性能有所改善.

荚醚类萃取剂用于分离HLLW是近年来的研究热点.中国原子能科学研究院等研究单位在过去20年研究了多种酰胺荚醚(TODGA)萃取剂对An的萃取化学行为,并用筛选出的荚醚萃取剂开展了一系列组分离实验,用含N类软配体萃取剂开展了An(Ⅲ)和Ln(Ⅲ)的分离研究.同时,开发了基于TODGA提取Sr的流程.国内其他研究单位还进行了将分离试剂与固相材料相结合的固液吸附体系分离An、提取Sr的研究.本文总结了这方面的研究进展,并提出了进一步研究建议.

以N,N,N',N'-四辛基-3-氧戊二酰胺(TODGA)为代表的酰胺荚醚类萃取剂可以有效萃取高放废液中的An(Ⅲ)和Ln(Ⅲ),为防止Zr4+、Pd2+等裂片元素萃入有机相,通常需要加入H 2 C2 O4作为水相络合剂,目前,H 2 C2 O4对TODGA萃取Ln(Ⅲ)的影响尚未报道.本工作研究了HNO3、H 2 C2 O4浓度对TODGA或TODGA+TBP体系萃取Nd3+的影响,同时测定了有机相中的H 2 C2 O4浓度,并用紫外-可见吸收光谱分析了有机相中的H 2 C2 O4与有机相中Nd3+的配位情况.研究结果表明:HNO3浓度在1.0～3.0 mol/L的范围内,Nd3+的分配比D(Nd3+)随HNO3浓度的增加而增加;H 2 C2 O4浓度在0.1～0.5 mol/L的范围内,D(Nd3+)随H 2 C2 O4浓度的增加而增加.HNO3浓度在1.0～3.0 mol/L的范围内,萃入有机相中H 2 C2 O4浓度随HNO3浓度的增加而减小,且存在于有机相中的H 2 C2 O4并未与有机相Nd3+配位.

合成了N,N′-二乙基-N,N′-二苯基-[2,2′-联吡啶]-6,6′-二硫代酰胺(Et-Ph-BCTABipy)萃取剂,并利用13 C N M R和1 H N M R对其进行了表征;研究了相接触时间、萃取剂浓度、水相初始酸度和Pd2+浓度等因素对Et-Ph-BCTABipy萃取Pd2+性能的影响,利用摩尔比法确定了Et-Ph-BCTABipy与Pd2+所形成的配合物组成;同时,在Ln(Ⅲ)与Pd2+共存体系中研究了Et-Ph-BCTABipy对Pd2+的萃取选择性.结果表明:Et-Ph-BCTABipy在HNO3体系中对Pd2+具有较强的萃取性能和较高的萃取选择性;萃取过程中Et-Ph-BCTABipy与Pd2+以1:2的比例结合,其萃取平衡常数Kex=3.42×106.

为了模拟Pu4+水解聚合行为,采用动态光散射法研究了盐酸溶液中Fe3+水解聚合边界条件,并采用热力学软件HSC Chemistry 6.0模拟计算了水解聚合过程中一些热力学函数值的变化.结果表明:在Fe3+发生水解聚合反应时,溶液中的c0(H+)、c(Fe3+)和温度之间存在一定的关联性,提高Fe3+浓度和升高温度会促进Fe3+水解聚合反应;聚合胶体粒径随着Fe3+浓度和温度升高而分布越广,且测得中值粒径在100～1000 nm.

以N,N,N',N'-四辛基-2-甲基-3-氧戊二酰胺(Me-TODGA)或N,N,N',N'四辛基-3-氧戊二酰胺(TODGA)为萃取剂、磷酸三丁酯(TBP)为相改良剂、煤油为稀释剂,对比研究了水相酸度、萃取剂浓度、锶浓度、温度对Me-TODGA-TBP体系和TODGA-TBP体系萃取Sr2的影响,并采用斜率法确定了萃合物的组成.结果 表明,2种酰胺荚醚萃取Sr2+的分配比(Dsr)随HNO3浓度(c(HNO3) =0.1～2.7 mol/L)、萃取剂浓度(c(萃取剂)=0.05～0.3 mol/L)的增加而增大,随Sr2+浓度的升高略有下降,随温度的升高而下降.2种萃取剂的萃合物组成分别为Sr(NO3)2·3Me-TODGA和Sr(NO3)2·2TODGA.萃取反应的△H分别为-69.46 kJ/mol和-51.39 kJ/mol,△S分别为-190.5 J/(mol·K)和-128.4 J/(mol·K),△G分别为-12.68 kJ/mol和-13.12 kJ/mol.相比之下,Me-TODGA萃取Sr2+的分配比不到TODGA的1/5.

通过采用密度泛函理论研究Pu3+、Pu4+、PuO2+和PuO22+离子与水分子结合形成水解机理.对比Pu3+、Pu4+、PuO2+和PuO22+的水合能,发现当结合相同数目的水分子时pu4+的水合物最稳定.此外,发现相比非相对论密度泛函理论计算,Pu-O配位键的键长约缩短10％.但是,计算的结合能与非相对论近似结果差异不大,这是由于计算结合能时相对论效应部分被抵消.理论研究Pu3+、Pu4+、PuO2+和PuO22+离子的水解过程,发现存在两种水解方式,分析吉布斯自由能变化,发现Pu离子直接与溶液中的氢氧自由基进行配合较易发生.而对于Pu4+离子最优的水解方式为夺取水中的氢氧根而产生水合氢离子.采用分子动力学模拟(MD),发现pu3+离子与pu4+离子分别倾向与9个水分子和8个水分子结合,PuO2+离子和PuO22+离子与5个水分子结合时最稳定.因为MD考虑了周期性边界条件,我们认为其相比密度泛函理论计算结果更可靠.

Background: Extraction and separation of neptunium is one of the focused research topics in the field of reprocessing. As an organic salting-free reductant, because of its reduction rate differences between Np(Ⅵ) and Pu(Ⅳ), methyl hydrazine (MMH) has the potential to separate Np/Pu.Purpose: In this study, the feasibility of applying MMH to reductive stripping separate Np/Pu was evaluated.Methods: The reductive back-extraction dynamics of Np(Ⅵ) and Pu(Ⅳ) by MMH were investigated using a single-stage extraction device.Results: By investigating the effects of the concentration for nitric acid and methyl hydrazine nitrate, as well as reaction temperature on the reductive stripping process, the Np(Ⅵ) and Pu(Ⅳ) stripping kinetics equation and their apparent activation energy were determined.Conclusion: Further, the half reaction time of reductive stripping Np(Ⅵ) and Pu(Ⅳ) with hydrazine nitrate can be obtained by means of kinetics equation, and a preliminary exploration for Np(Ⅵ)/Pu(Ⅳ) separation was carried out.

镎的提取和分离是国际后处理领域重点关注的研究课题之一.在Puex流程中,硝酸肼常被用来作为亚硝酸的清扫剂,此外,由于硝酸肼对Np(Ⅳ)和Pu(Ⅳ)的氧化还原反应具有选择性,理论上可以利用其反应速率上的差异来实现镎与铀钚的分离.为探索硝酸肼分离镎/钚工艺提供可行性,本文采用单级萃取设备研究了硝酸肼还原反萃Np和Pu的过程.通过研究硝酸浓度、硝酸肼浓度和反应温度对还原反萃过程的影响,确定了Np(Ⅵ)和Pu(Ⅳ)反萃动力学方程和表现活化能.进一步通过动力学方程得出硝酸肼还原反萃Np(Ⅵ)和Pu(Ⅳ)的半反应时间,并对Np(Ⅵi)/Pu(Ⅳ)分离过程的工艺进行了初步探索.

In order to understand the chemical behavior of neptunium during the extraction process,the oxidation behavior of Np(Ⅴ)to Np(Ⅵ)in nitric acid solution and the extrac-tion performance of Np(Ⅵ)by 30%TBP/kerosene was studied by single stage extraction. The experiment results indicate that the increasing of the concentrations of nitric acid and nitrous acid are in favor of the production of Np(Ⅵ).The reaction rate of Np(Ⅴ)to Np(Ⅵ) by oxidation increases with increasing the reaction temperature.The simulated experiment on the mixer-setter set indicates that the extraction ratio of neptunium in 1 A extraction section is approximately 80% when the HNO2 concentration in 1AX is 0.01 mol/L,the extraction temperature is 45 ℃ and the HNO3 concentration in 1AF is 3.5 mol/L HNO3 .

It was found that in the nitric acid solution containing U (Ⅳ ) ,hydrazine , Tc(Ⅶ) and Np(Ⅴ) ,Np(Ⅴ) was reduced to Np(Ⅳ ) quickly .The main reason of the fast reduction of Np(Ⅴ) was due to Tc(Ⅳ ) ,which was produced by the reduction of Tc(Ⅶ) in the presence of U (Ⅳ) .The routing of Np in U/Pu separation stage was stud‐ied by tube cascade experiments and bench scale experiments at a miniature mixing‐settler set .The results show that nearly 67 .7% of Np goes into 1BU in the presence of Tc in the 1AP ,but approximately 27 .0% of Np goes into the 1BU in the absence of Tc . The amount of Np going into 1BU increases with the reaction temperature and the concentration of Tc(Ⅶ) in 1AP .

The process of Pu reductive stripping by hydroxylamine nitrate in plutonium purification cycle was optimized by the single extraction experiments.Hydrazine nitrate can reduce a small quantity of Pu(Ⅳ)and strip it from 30% TBP-kerosene into an aqueous phase,but can suppress the plutonium reductive stripping with increasing hydrazine nitrate concentration,where hydrazine nitrate plays a role as a salting-out agent.For the Pu purification cycle,the Pu recovery can be improved by enhancing the reductant flow when keep the HAN/Pu molar ratios invariable.2-3 is feasible for the HAN/Pu molar ratios when the temperature is 50 ℃.Pu concentration in organic phase increases with raising the reaction temperature because reaction rate of HNO3 with Pu(Ⅲ)is accelerated,and also increases with increasing ionic strength because the distribution coefficient of Pu(Ⅲ)is amplified due to the salting-out effect.

The reduction of NpO2+ to Np4+by U(Ⅳ)in nitric acid solution has been studied by spectrophotometric method,and the rate equation is described by:-dc(Np(Ⅴ))/dt=kc(Np(Ⅴ))c0.7(U(Ⅳ))c1.9(H+)c(NO-3) where k=(6.37±0.49)×10-3 L3.6/(mol 3.6·min)at 25℃.The activation energy is estimated to be 60.13kJ/mol.U(Ⅵ)with the concentration of 0-4.2×10-2 mol/L has little effect on the reaction.The possible mechanism is discussed.

The development and its corresponding technical features of spent nuclear fuel reprocessing were reviewed systematically according to the changes of its applications to different spent fuels and separation improvements. Aiming at the partition and transmutation (P&T) technologies in future advanced nuclear energy system, the improvements of the Purex process from Generation II reprocessing to Generations III and IV reprocessing were highlighted. The key radiochemical issues which should pay much attention in the Purex process and following partition processes as well as the dry reprocessing for spent nuclear fuel of fast reactors were summarized.