镅锔分离研究进展

乏燃料后处理产生的高放废液中Am和Cm是长期释热的主要来源,将它们分离出来并进一步进行分离和处置,对高放废物的长期安全处理处置具有重要意义。另外,超钚元素生产涉及Am和Cm材料的获取以及辐照后靶件中Am和Cm的化学分离。因此Am、Cm的分离一直是锕系元素化学与材料研究的重要领域之一。但是Am、Cm之间的分离相当困难,水溶液中Am、Cm基本均以正三价离子形式存在,化学性质非常相似。早期的离子交换法分离因子低,近年来主要研究将Am(Ⅲ)氧化到高价态实现分离,或通过Am、Cm与配体的亲和力差异、不同配体组合产生“推拉效应”以提高分离因子。本文综述了相关研究现状,概述了主要流程研发情况,并展望了该领域的研究趋势。     

Neutron Cross Sections of Li-6

Calculation procedures have been developed to evaluate the performance of the multistage counter current extraction of transuranics (TRU) from spent molten salt into liquid metal, taking into account stage efficiency and also the scrub stage. The following results, which supplement previous papers, were derived using these procedures. When Cd is used as the liquid metal and the stage efficiency is assumed to be 100%, at least four stages are necessary to recover 99% of TRU from the salt with a decontamination factor (DF) higher than five. A stage efficiency of the extraction better than 80% is desirable for a practical application. The scrub stage is not very effective in improving the DF when the total number of extractions is less than five. The DF slightly increases with higher TRU concentration in the salt since the accompanying lanthanide FP extracted into the Cd in the later stages works as a mild reducing agent in the earlier stages. Although the extraction process has high separation capability, it is very difficult to separate Np, Am, or Cm from Pu due to their similar separation factors. Therefore, the extraction process has inherent proliferation resistance.     

Extraction of Some Elements by Mixture of DIDPA-TBP and Its Application to Actinoid Partitioning Process

Effect of TBP on the extraction of Pu(IV), Zr(IV), Nb(V), Nd(III) and Am (III) was studied with diisodecylphosphoric acid(DIDPA) as an extractant. Hydrolyzable elements contained in the high-level liquid waste of fuel reprocessing,i.e. Pu(IV), Zr(IV) and Nb(V), could be extracted with mixtures of DIDPA and TBP of different compositions. Addition of TBP to DIDPA causes an increase and a decrease of respectively distribution rate and ratio of Zr(IV). These elements extracted were completely stripped with the aid of oxalic acid. An effect of TBP on separation of transplutonides (III) from lanthanoids(III) in the DIDPA and DTPA extraction system was also studied.

Based on the results, a process flow sheet utilizing the extractant of DIDPA and TBP mixture was contrived for partitioning actinoids in the high-level liquid waste.     

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.     

Conversion of Simulated High-level Liquid Waste to Chloride for the Pretreatment of Pyrometallurgical Partitioning Process

A pyrometallurgical partitioning process is being developed for recovering minor actinides from high-level liquid waste resulting from PUREX reprocessing. Since the high-level liquid waste consists of concentrated raffinate, concentrated alkaline waste and insoluble residues, the various elements in the waste must be converted to chlorides before they can be sent on to the pyrometallurgical partitioning process. The conversion to chlorides is done by a combination of denitration and chlorination. The mass balance of these processes was measured in the present study using simulated high-level liquid waste. The results indicate that almost all of the alkali elements and Re, substituting for Tc, and significant amounts of Se, Cr, and Mo were separated by denitration, and that Cr, Fe, Zr, Mo, and Te were separated by chlorination. The remaining noble metals, Ni, U, and alkaline-earth and rare-earth elements were efficiently converted to chlorides, which were then supplied to the reductive extraction test using a molten salt/liquid-Cd system to demonstrate that the obtained chlorides are appropriate for processing by pyrometallurgical partitioning. In further reduction, noble metals and Ni were reductively extracted into the liquid-Cd phase, and the rare-earth elements and U into the liquid-Cd phase by adding Li reductant. These elements were completely separated from the alkaline-earth elements remaining in the chloride phase.     

Solvent Extraction of Americium(VI) by Tri-n-Butyl Phosphate

Solvent extraction of Am(VI) by tri-n-butyl phosphate (TBP) from nitric acid solutions was investigated to develop a novel method for partitioning americium from high level liquid waste generated through spent nuclear fuel reprocessing. Am(VI) was prepared using ammonium peroxodisulfate and silver nitrate. The distribution coefficients of Am(VI) were determined for extraction systems of various concentrations of nitric acid and TBP. Sufficiently stable Am(VI) could be extracted and the extraction reaction of Am(VI) was found to be the same as for other hexavalent actinides. The apparent equilibrium constant varied with the concentration of peroxodisulfate used for the valence control, which was ascribed to the competitive reaction of the extraction of Am(VI) and the complex formation of Am(VI) with sulfate ion produced by the decomposition of peroxodisulfate. A distribution coefficient of Am(VI) above 1 was obtained with undiluted TBP and the separation factor between Am(VI) and Nd(III) was 87±9. TBP extraction of Am(VI), after implementing valence control, was proved to be an effective method for the partitioning of americium from fission products such as rare earth elements.     

Wide Range Wave-Number Spectra of Place-to-Place Variations in Terrestrial Gamma-Ray Exposure Rate

The precipitation behavior of Pu. Np and Am during the denitration of high-level radioactive liquid waste (HLW) by formic acid was studied using a simulated HLW. The dissolution of the precipitate formed in denitrated HLW was also studied using oxalic acid to recover transuranium (TRU) elements from the precipitate. In the denitration, the precipitated fractions of TRU elements increased with decreasing acidity of the denitrated HLW. In the denitration at [HCOOH]/[HNO₃]=1.5, which was adopted in the partitioning process developed in JAERI, the precipitated fractions of Np and Am were only 0.6% and 0.06%, respectively, whereas that of Pu was 90 %. The precipitation fractions of Pu and Np did not depend on their concentrations in the range of 6x10^?5–6x10^?4 M for Pu and 10^?5–10^?3 M for Np. Plutonium was not precipitated itself by polymerization or hydrolysis but coprecipitated with other elements such as Mo and Zr. It was found that the precipitate formed during the denitration of 1 l of HLW could be dissolved in a 800 ml of 0.5 M oxalic acid solution.     

Separation of Actinide Elements by Solvent Extraction Using Centrifugal Contactors in the NEXT Process

Using the advanced aqueous reprocessing system named NEXT process, actinides recovery was attempted by both a simplified solvent extraction process using TBP as an extractant for U, Pu and Np co-recovery and the SETFICS process for Am and Cm recovery from the raffinate. In U, Pu and Np co-recovery experiments a single cycle flow sheet was used under high nitric acid concentration in the feed solution or scrubbing solution. High nitric acid concentration in the feed solution aided Np oxidation not only in the feed solution, but also at the extraction section. This oxidation reaction accomplished Np extraction by TBP with U and Pu. Most of Np could be recovered into the product solution. In the SETFICS process, a TRUEX solvent of 0.2 mol/dm³ CMPO and 1.4 mol/dm³ TBP in n-dodecane was employed instead of 0.2 mol/dm³ CMPO and 1.0 mol/dm³ TBP in n-dodecane in order to increase the loading of metals. Instead of sodium nitrate, hydroxylamine nitrate was applied to this experimental flow sheet in accordance with a “salt-free” concept. The counter current experiment succeeded with the Am and Cm product. On the high-loading flow sheet, compared with the previous flow sheet, the flow of the aqueous effluents and spent solvent were expected to decrease by about one half. Two solvent extraction experiments for actinides recovery demonstrated the utility of the flow sheet of these processes in the NEXT process.     

Controlling Neptunium and Plutonium within Single Cycle Solvent Extraction Flowsheets for Advanced Fuel Cycles

Simple hydroxamic acids are shown to be useful reagents for the separation of Np and Pu from U within simplified, single cycle Purex flowsheets. They are compatible with the use of centrifugal contactors and laboratory scale flowsheet trials with aceto-hydroxamic acid have demonstrated high actinide recoveries and decontamination factors on products for active feeds of up to 40 wt.% Pu. They therefore show many ideal characteristics for Pu and Np recovery within flowsheet options for actinide recovery in advanced fuel cycles. Furthermore, in order to optimize the routing of Np with the Pu product in advanced flowsheets, additional studies of Np extraction in the primary co-decontamination contactor, prior to U/Pu partition, have been undertaken, combining experiment, modelling and flowsheet tests.     

二(2,4,4三甲基戊基)-二硫代膦酸萃取分离Am和Cm的研究

研究了二（2,4,4三甲基戊基）-二硫代膦酸（HBTMPDTP）对Am与Cm的萃取和分离,单级萃取分离因数约为3。pH值、离子强度、萃取剂浓度、温度等因素对分离因子影响不大。8级萃取、3级洗涤的多级逆流萃取实验表明：HBTMPDTP能够使Am与Cm得到有效分离。萃取实验的计算值和实验值符合很好。根据串级计算给出了HBTMPDTP萃取分离压水堆废液中镅和锔的推荐流程,采用9级萃取、4级洗涤的萃取分离流程,镅的萃取率为99．92％,放化纯度达到99．99％,质量纯度达到99．82%;锔的萃余率为94．90％,放化纯度达到99．90％,质量纯度达到97．68％,镅中锔的分离因数为20,锔中镅的分离因数为1186。可以满足Am与Cm的分离-嬗变要求。     

用TRPO从高放废液中去除锕系元素的研究──微型离心萃取器串级冷实验

采用微型离心萃取器进行了ＴＲＰＯ流程从模拟高放废液中去除锕系元素的冷实验。实验中用Ｎｄ代替Ａｍ，Ｚｒ代替Ｎｐ、Ｐｕ，在模拟高放废液稀释３倍、酸度为１．０ｍｏｌ／ｌ时，采用１２级萃取、４级洗涤能有效地去除模拟高放废液中９９．９％以上的Ｎｄ、Ｚｒ，满足了冷实验要求，并且萃取中不出现三相，可以使萃入的Ｆｅ洗下６０％，避免大量Ｆｅ进入后续流程。采用硝酸、草酸分别反萃Ｎｄ和Ｚｒ，使Ｎｄ和Ｚｒ分成二组，交叉污染很小。文中给出了硝酸、Ｎｄ、Ｚｒ等在各级的浓度剖面和它们在各物流中的分布。     

Development of a simplified separation process of trivalent minor actinides from fission products using novel R-BTP/SiO2-P adsorbents

From a viewpoint of direct separation of trivalent minor actinides (MA: Am, Cm etc.) from fission products (FP) including rare earths (RE) in high level radioactive liquid waste, the authors have developed a simplified separation process using a single column packed with novel extraction adsorbents. Attention was paid to a new type of nitrogen-donor ligand, R-BTP (2,6-bis(5,6-dialkyl-1,2,4-triazin-3-yl)pyridine, R: alkyl group) as an extractant because it has higher extraction selectivity for Am(III) than RE(III). Since the R-BTP ligands show different properties such as adsorbability and stability when they have different alkyl groups, several R-BTP extraction adsorbents were prepared by impregnating the R-BTP ligands with different alkyl groups (isohexyl-, isoheptyl- and cyheptyl-BTP) into a porous silica/polymer composite support (SiO₂-P particles). This work investigated: (1) fundamental properties of the synthesized R-BTP/SiO₂-P adsorbents, (2) adsorption and desorption properties of Am and FP in nitric acid solution and water using the adsorbents in a batch experiment, (3) radiolytic and chemical stabilities of the adsorbents, and (4) the possibility for developing a simplified separation process of MA using the most promising adsorbent (isohexyl-BTP/SiO₂-P) under temperature control between 25 and 50°C.     

Continuous extraction and separation of Am(III) and Cm(III) using a highly practical diamide amine extractant

A highly practical diamide-type extractant, which is an alkyl diamide amine with 2-ethylhexyl alkyl chains (ADAAM(EH)), was investigated for the mutual separation of Am(III) and Cm(III). ADAAM(EH) is a multidentate ligand with one soft N-donor atom and two hard O-donor atoms as part of its central frame. This tridentate arrangement of donor atoms provides selective binding to Am(III) compared to that with Cm(III) in highly acidic media (1.5 M HNO₃), resulting in separation factors of up to 5.5. A continuous liquid–liquid extraction and stripping test was conducted using a multistage countercurrent mixer-settler extractor with ADAAM(EH) in n-dodecane. In this test, the separation of Am(III) and Cm(III) was achieved with very high yield.     

Separation of Curium from Neutron Irradiated Am-241

The conditions under which curium can be separated from irradiated ²⁴¹Am target were elucidated. The isolation process consists of three steps: In the first step, Am(III) is oxidized to pentavalent state in a dilute nitric acid solution, and then plutonium and curium are extracted from the irradiated target by solvent extraction with HDEHP. Curium in the organic phase is back-extracted with 1 N nitric acid, and thereafter plutonium with a reducing solution containing ferrous sulfamate. The curium is finally purified by cation exchange, using α-hydroxy isobutyrate as the eluting solution. About 0.4 μg of ²⁴²Cm and 4×10^?3 μg of ²⁴³Cm were found in the curium fraction, which had been separated from 1 mg of irradiated ²⁴¹Am sample.     

Separation of Transplutonium and Rare-Earth Elements by Extraction with Di-isodecyl Phosphoric Acid from DTPA Solution

The distribution ratio Df and the separation factor β for Nd(III) to Am(III) were studied in DIDPA-DTPA systems to determine optimum conditions for applying DIDPA to the TALSPEAK type extraction process to separate transplutonides from lanthanoids in the partitioning of high-level waste of nuclear fuel reprocessing.

Extraction of lanthanoids from 0.05–0.1 M DTPA-1 M lactic acid (pH 3.0) aqueous solution into 0.2–0.3 M DIDPA in DIPB gives the separation factor of 20 revealing practicability of this system in the partitioning. The nature of diluent affects greatly Df, and DIPB proved to be the most appropriate one for the separation of transplutonides from lanthanoids. The presence of lactic acid in the aqueous phase improved the extraction kinetics in DIDPA-DTPA system.     

乙羟肟酸消除Zr界面污物的研究

研究在模拟高放废液中加入乙羟肟酸（AHA）以消除酰胺荚醚（TBOPDA）萃取模拟高放废液过程中的界面污物。萃取实验结果表明：在模拟高放废液中加入AHA可显著降低Zr（Ⅳ）在两相中的分配比，此时，Pu（Ⅳ）的分配比仍足够大，它不影响TBOPDA对Pu（Ⅳ）的回收。反萃实验表明：在所研究的反萃条件下，1级反萃即可有效反萃TBOPDA有机相中的Zr（Ⅳ）；3次错流反萃可有效反萃TBOPDA有机相中的Pu（Ⅳ）；反萃液中加入AHA对Am（Ⅲ）的累计反萃率影响很小；提高反萃液的酸度可抑制TBOPDA有机相中Am（Ⅲ）的反萃。     

Continuous separation of molybdenum and zirconium from simulated high-level liquid waste with a Taylor–Couette contactor

Reprocessing of spent nuclear fuels generates high-level liquid waste (HLLW) which undergoes vitrification into borosilicate glass before final geological disposal. To ensure the quality of the glass, control of the concentration of chemical species such as molybdenum (Mo), which has an adverse impact on the vitrification process, is critical. Also, zirconium (Zr) can cause crud in washing process and Zr-93 is a long-lived fission product needed to be separated. In this study, a liquid–liquid countercurrent centrifugal contactor with Taylor–Couette flow (TC contactor) was applied to practical multi-species cases. Continuous separation of Mo and Zr from a simulated HLLW with bis(2-ethylhexyl) phosphoric acid (HDEHP) as extractant has been performed. Among a variety of metals in simulated HLLW, Mo, Zr, Y, and Fe are extractable, Mo and Zr were separated from HLLW by equilibrium, and Fe/Y separation was achieved by the effect of non-equilibrium state in TC contactor. Addition of tributyl phosphate could improve extraction of Mo. This study has expanded the scope of the TC contactor to multi-species separation processes.     

动力堆模拟高放废液的甲醛脱硝及脱硝过程中的沉淀行为北大核心CSCD

两步法玻璃固化工艺中,高放废液可通过化学脱硝达到降低酸度的目的,常用的化学脱硝剂有甲酸、甲醛、蔗糖等。以甲醛为化学脱硝剂,对动力堆模拟高放废液进行脱硝及脱硝过程中沉淀行为进行研究。模拟高放废液在90℃、脱硝比例为1.0~2.0范围内进行脱硝,对脱硝后各物质运用电感耦合等离子发射光谱(ICP-OES)、X射线荧光光谱(XRF)、扫描电镜-能谱分析(SEM-EDS)、Raman光谱、X射线衍射(XRD)、热重分析仪(TG)进行分析。结果表明:脱硝后废液中NO_(3)^(-)含量明显降低,随脱硝比例增大,NO_(3)^(-)的含量逐渐降低,甲醛含量增加。脱硝过程中出现由Zr、Mo、La、Ce、Nd、Fe、Te、Pr、Cs、Sm、Cr、Sr、Y、Co、Ni组成的沉淀,沉淀的形成有两个过程:一个过程为形成颗粒状的结晶物Ln_(2)Zr_(3)(MoO_(4))_(9)(Ln=La、Ce、Nd、Pr、Eu、Sm)和MoO_(2);另外一个过程为形成由O、Fe、Zr、Mo、Te构成的无定形粉末物质;脱硝产物的热分解主要发生在约360℃以下。     

Heterogeneous Transmutation of Am, Cm, and Np in a BREST Core

The effect of heterogeneous (in individual units) transmutation of Am, Cm, and Np on the radiation characteristics of fuel is examined. For heterogeneous transmutation of Am, Cm, and Np in individual fuel elements containing nitrides, the radiation characteristics and energy release increase substantially compared with fresh homogeneous fuel elements. Heterogeneous transmutation is dangerous from the standpoint of nonproliferation of fissioning materials because of the low critical mass of the main nuclides – ²³⁹Np and americium isotopes. 2 tables, 4 references.     

Actinide reformer concept

The partitioning and transmutation technology is effective to reduce the environmental impact from disposition of high-level radioactive wastes and improve the efficiency of geological disposal. However, Am and Cm and their daughter nuclides are difficult to handle in the fuel cycle because of their high decay heat and radioactivity. These nuclides also give the chemical instability which harms the soundness of fuel pellet properties.

We propose a new system concept “actinide reformer”, which reforms Am and Cm into Pu by neutron capture reactions and decay. Am and Cm are separated from the PUREX reprocessing process and converted to chloride molten-salt fuel. Using liquid-type fuel, above mentioned defects can be compensated. Actinide reformer is an accelerator-driven system which is composed of a 10 MW-class cyclotron, a tungsten target and a subcritical core. Spent molten-salt fuel is reprocessed as an on-line fuel exchange manner to extract fission products and recover Pu to send back to a power generation cycle. The decay heat and neutron emission from the fuel with recovered Pu are smaller than those of MOX fuel with 5% of minor actinide addition. It expects no significant engineering difficulties and cost increase for construction of MOX fuel based reprocessing/fabrication plant and power reactors.