用于锕系元素提取分离的萃取剂——TiAP

对磷酸三异戊酯(TiAP)和磷酸三丁酯(TBP)萃取体系的物理性质、萃取能力、耐辐照等方面进行了比较,结果表明,TiAP作为萃取剂在物理性质、萃取Pu(Ⅳ)和Np(Ⅳ)的能力以及辐照稳定性能等方面明显好于TBP。提出TiAP是一种很好的并有可能用于锕系元素提取分离的萃取剂。     

Extraction of Lanthanoids (III) with Di-isodecyl Phosphoric Acid from Nitric Acid Solution

The extraction of Nd(III) with di-isodecylphosphoric acid was studied and compared with that by di(2-ethylhexyl) phosphoric acid. DIDPA possesses higher D_f values in a strong acid medium but poorer separation factors against lanthanoids(III) than that of DEHPA. Mono (2-ethylhexyl) phosphoric acid increased extraction of Nd(III) with DIDPA in a similar manner as in the case of DEHPA, by a factor of D_f of 10³.

The extraction of Sr(II) was also investigated with DIDPA, which showed a maximum of D_f at pH 5.0. Results showed that DIDPA is a promising extractant for the partitioning of the high level liquid waste.     

Extraction of Pentavalent Neptunium with Di-Isodecyl Phosphoric Acid

As a fundamental study for separating neptunium from high-level radioactive waste, the mechanism and the rate of extraction of pentavalent neptunium with DIDPA were investigated. From the analysis of oxidation state of neptunium in organic phase, it was proved that disproportionation reaction of Np(V) was concerned in the extraction.

The reaction order of the extraction with respect to neptunium concentration was larger than unity. The extraction rate was much reduced by the decrease of DIDPA concentration. The dependence of the rate on nitric acid concentration and on the temperature was also examined.     

Pressure dependence of extraction behavior of plutonium(IV) and uranium(VI) from nitric acid solution to supercritical carbon dioxide containing tributylphosphate

Plutonium(IV) and uranium(VI) were extracted into supercritical CO₂ fluid phase (SF-CO₂) containing tributylphosphate (TBP) with equilibrium distribution ratios, D, e. g., D_Pu(IV) = 3.1 and D_U(IV) = 2.0, for the extraction of 2 × 10⁻³ M Pu(IV) and U(VI) from 3 M HNO₃ into SF-CO₂ containing 0.3 M TBP at 60 °C and 15 MPa. A simple linear relation between D and density of CO₂; long D = a log + b (a,b; constants), was observed, which was explained theoretically by the formulation of the extraction equilibrium taking into account the phase distribution behavior of extractant TBP and extracted species, i.e. Pu(IV)- and U(VI)-TBP complexes involved in the extraction reaction. The slopes a of the log D vs. log plots were −(1.6 ± 0.1) and −(2.7 ± 0.5) for the extraction of Pu(IV) and U(VI), respectively. The differences in D as well as the slope a between Pu(IV) and U(IV) make it possible to design the U and Pu separation method by which one can achieve an enhancement of the extraction efficiency and selectivity by tuning the operation pressure.     

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.     

Comparison of Four Bidentate Phosphoric and Diamide Compounds for the Extractability of Actinides

Using four bidentate phosphonate and diamide ligands (L), the extraction of Am(III), Th(IV), Np(V) and U(VI) was studied and their extraction species and extractability in these extractants were discussed. The extractant containing the phosphoric oxygen has stronger chelation than that with the carbonyl oxygen atoms. The molar ratios in the main chemical forms extracted are 1:3 for Am:L, 1:1 for Th:L and Np:L and 1:2 for U:L from the relationship between log D(An) and log[L]. The order of D in actinides with the different oxidation states is U(VI)>Th(IV)> Am(III)>Np(V), which shows the different order from that for HDEHP and HTTA, namely Th(IV)>U(VI)> Am(III)>Np(V). It is clear that BDPPM, a diphosphoric dioxide extractant, has high extractability towards all An used here.     

Selective uptake properties of metal ions by hybrid microcapsules enclosed with TBP

The granulation of TBP extractant is effective for the enhancement of uptake efficiency. The granulation was accomplished by microencapsulating techniques using alginate gel polymers (alginate and alginic acid gel polymers; calcium alginate, barium alginate and nitric alginate (CaALG, BaALG and HALG)). The characterization of hybrid microcapsules was examined by SEM/EPMA, and the uptake properties and the selectivity of various nuclides, Fe(III), Sr(II), Co(II), U(VI) and Pu(IV), were examined by batch methods. A relatively high uptake (%) of Fe(III), Sr(II) and Co(II) above 80% was obtained in the presence of 10⁻³ M HNO₃, and the uptake equilibrium was attained within 5 h. The uptake rate of U(VI) and Pu(IV) attained equilibrium within 1 h and 3 h, respectively. At higher HNO₃ concentration ranging from 10⁻³ M to 5 M, the uptake (%) of Fe(III), Sr(II) and Co(II) was considerably lowered. In contrast, the uptake (%) of U(VI) and Pu(IV) about 60% was obtained even in the presence of 5 M HNO₃. The uptake of U(VI) for MCs (TBP–CaALG) was governed by the extraction with TBP micro droplets and ion-exchange reaction in the CaALG matrices. Energy dispersive spectra (EDS) showed that U(VI) ions were incorporated into both phases of TBP and CaALG in microcapsules.     

Application of N,N-di(2-ethylhexyl)butanamide for mutual separation of U(VI) and Pu(IV) by continuous counter-current extraction with mixer-settler extractors

Continuous counter-current extraction using N,N-di(2-ethylhexyl)butanamide (DEHBA) as an extractant was performed with mixer-settler type extractors consisting of U–Pu extraction, scrub, U recovery, Pu back-extraction, and U back-extraction steps. The feed solution used in the continuous counter-current extraction was 3 mol/dm³ (M) nitric acid containing U, Pu, and simulated fission products of Sr, Ba, Zr, Mo, Ru, Rh, Pd, and Nd. More than 99.9% of U and Pu in the feed was extracted by 1.9 M DEHBA at the U–Pu extraction step with negligible extraction of Sr, Ba, Mo, Ru, Rh, and Nd. The extracted Pu was back-extracted via contact with 0.3 M nitric acid in the Pu back-extraction step, and the ratio of Pu distributed to the Pu fraction stream was ～ 82%. It was confirmed that 1.9 M DEHBA effectively recovered U in the U recovery step, and the ratio of U in the Pu fraction stream was less than 1%. The extracted U was back-extracted in the U back-extraction step, and more than 98% of U was recovered in the U fraction stream.     

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.     

Fractional Extraction of Zirconium,Niobium and Neptunium with Tri-n-butyl Phosphate

Radioactive Zr, Nb and Np were separated from each other by a continuous fractional extraction method.

The extraction of Zr and Nb was carried out with 1 % TBP in chloroform from the 10N hydrochloric acid solution of irradiated U. Under these conditions, Zr, Nb and Np could be extracted continuously with the organic solution, while U and other F.P. remained in the aqueous solution. Finally, Zr in the organic solution was back-extracted with concentrated hydrochloric acid containing 1sim;2 drops of hydrofluoric acid.

The method is convenient for the separation of Zr, Nb and Np from the F.P.

The distribution ratios in the extraction system have also been obtained for a number of radioactive nuclides.     

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.     

萃取分离法处理高放废液的进展

评述了近几年用萃取分离法从高放废液中去除超铀锕系元素的进展情况，着重介绍世界上已有的应用前景较好的ＴＲＵＥＸ流程（美）、ＤＩＡＭＥＸ流程（法）、ＤＩＤＰＡ流程（日）、ＣＴＨ流程（瑞典）和ＴＲＰＯ流程（中国）。     

H2C2O4对TODGA萃取Nd3+的影响

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H₂C₂O₄作为水相络合剂,目前,H₂C₂O₄对TODGA萃取Ln(Ⅲ)的影响尚未报道。本工作研究了HNO₃、H₂C₂O₄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H₂C₂O₄浓度,并用紫外-可见吸收光谱分析了有机相中的H₂C₂O₄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;H₂C₂O₄浓度在0.1~0.5 mol/L的范围内,D(Nd³⁺)随H₂C₂O₄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H₂C₂O₄浓度随HNO₃浓度的增加而减小,且存在于有机相中的H₂C₂O₄并未与有机相Nd³⁺配位。     

Modified TRUEX Process for the Treatment of High-Level Liquid Waste

The TRUEX process has been examined to recover Am and Cm from the high-level liquid waste of Purex reprocessing plant. Continuous counter current extraction and back-extraction experiments were carried out by a mixersettler using simulated waste solution for three process flowsheets, i.e. a process flowsheet developed by Argonne National Laboratory and other two process flowsheets which were modified in the scrub stage. The result indicates that the process flowsheet of Argonne National Laboratory cannot be applied for the high-level liquid waste containing high concentrations of lanthanide and actinide elements because of the formation of insoluble salts of these elements with oxalic acid, which is added to restrict the extraction of fission products such as Mo and Zr. A modified process flowsheet, which had only one scrub stage with high concentration of nitric acid, was found to be the best of three process flowsheets examined, where Nd as a simulated element of Am and Cm was sufficiently recovered and any precipitation of oxalate salt was not observed.     

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。     

Pulsed Neutron Experiment in Fast Natural Uranium Systems

Separation of fission products was carried out by solvent extraction using tri-n-butyl phosphate (TBP). ^{144, 144}Ce, ⁹¹Y and ⁹⁵Zr and ⁹⁵Nb were extracted with TBP from a freshly prepared nitric acid-potassium bromate solution. These nuclides in TBP were successively back-extracted with various aqueous solutions: ^{144, 144}Ce with hydrogen peroxide and nitric acid solution, ⁹¹Y with hydrochloric acid solution, and ⁹⁵Zr-⁹⁵Nb with oxalic acid solution. The other nuclides were not extracted by the TBP and remained in the nitric acid and potassium bromate solutions.     

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.     

Solubility of Tributylphosphate in Plutonium Nitrate Solution and Highly Radioactive Liquid Waste Solution

Abstract

The solubility of tri-n-butylphosphate (TBP) in aqueous solutions of plutonium nitrate (PuN) and in highly radioactive liquid waste (HRLW) of PUREX nuclear fuel reprocessing was investigated. By an empirical formula the solubility of TBP in PuN solutions was described in the range of 0–0. 1 M Pu and 1–8M HNO₃ concentrations. The following items were elucidated:

(1) The logarithm of TBP solubility (S) in the solution of interest varies inversely in proportion to the concentration of Pu(IV) in the range of 0–0.1M PU(IV) at a constant concentration of HNO₃, indicating that Pu(IV) simply behaves as an electrolyte for the salting-out of TBP. Log S subsequently levels off with increasing Pu concentration, which would be due to a change in the principal dissolution form of TBP having an interaction with Pu (IV).

(2) The variation in S in PuN solutions (0–0.1M PU) with nitric acid concentration shows almost the same tendency as that in HNO₃ solution.

(3) A dependency of S on fission product metal ions in HNO₃ for HRLW similar to that for PuN was observed.

(4) The logarithm of the ratio of TBP solubility in water to that in solution of interest was nearly proportional to l/T for HRLW solution or for low concentration of PuN solution. That deviates from the linear relation at high temperature when the concentration of PuN is increased, which can be explained by the change in ionic form of Pu.     

Separation of zirconium and hafnium by solvent extraction using mixture of TBP and Cyanex 923

In this study, a new Zr/Hf separation procedure by solvent extraction is proposed. The method is based on using the mixture of TBP and Cyanex 923 as extractant in the organic phase. Several parameters including; TBP/Cyanex 923 volume ratio, extractant concentration in the organic phase, nitric acid and NaNO₃ concentration in aqueous phase have been investigated. The results demonstrate better condition respect to traditional TBP/HNO₃ process. Lower required acid concentration and more zirconium extraction are some advantages of the new separation process.     

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.