Plasma separation of the elements applied to nuclear materials handling

Physical plasma method of separation by mass of the constituent elements together with chemical methods could find application in the fuel cycle of fast recators. These methods could also be useful for separating decay products during reprocessing of spent fuel of existing reactors and in the solution of questions concerning safety and nonproliferation. A possible plasma separation scheme is presented: solid-state material is transferred in a magnetic field into a low-temperature plasma flow and ions in a prescribed mass range are selectively accelerated and then spatially separated in a toroidal magnetic field. __________ Translated from Atomnaya énergiya, Vol. 101, No. 4, pp. 302–306, October, 2006.     

Separation of the alpha-emitting radioisotopes actinium-225 and bismuth-213 from thorium-229 using alpha recoil methods

An innovative method has been demonstrated for separating alpha-emitting isotopes for medical radiotherapy applications. The method relies on recoil-ion separation rather than on conventional wet chemistry techniques to separate medical isotopes from their precursor sources. The isotopes ²²⁵Ac and ²¹³Bi have been separated from electro-deposited sources of ²²⁹ThO₂. Separations of ²²⁵Ac were carried out by placing nickel recoil collector foils in firm contact with the ²²⁹ThO₂ sources. One-stage recoil-ion separations of ²²⁵Ac from ²²⁹Th have been performed as well as two-stage separations of ²¹³Bi from previously recoil separated ²²⁵Ac. In addition, a direct recoil separation of ²¹³Bi from ²²⁹Th has been demonstrated. The ²¹³Bi from the one-stage direct separation has a high isotopic purity, but contains small amounts of long-lived ²²⁵Ac alpha activity. The two-stage separations of ²¹³Bi produce high isotopic purity material (>99.9999%), but result in lower isotopic yields. Range-energy calculations have been carried out to determine the yields of recoil ions as a function of alpha-particle energy and ThO₂ thickness. The results of the calculations have been benchmarked with recoil separation measurements carried out using ThO₂ electro-deposits over a range of thickness. A computer code based on the generalized Bateman equations has been developed to allow calculations of the amounts of any isotope in the ²²⁹Th decay chain as a function of recoil separation exposure time and elapsed time after the separation. An excellent match has been obtained between the predictions of the Bateman calculations and the results of recoil separation measurements. The recoil separation method has proven to be a simple and effective way of separating medically useful isotopes such as ²¹³Bi. In addition, the method has been shown to produce no chemical or radioactive wastes, in contrast to radiochemical separation methods, which generate mixed (chemical and radioactive) waste.     

Uranium separation from a simulant fuel debris solution using a benzimidazole-type anion exchange resin

Ion exchange is one of the most useful separation methods. Certain ion exchange resins have been developed for use in reprocessing technology. The purpose of this study is to evaluate the U selectivity of a benzimidazole-type anion exchange resin and its potential for the separation of U from a solution of simulant fuel debris in HCl. The selectivity of the benzimidazole-type anion exchange resin for adsorption of the major elements present in fuel debris in HCl solution (such as in debris from the Fukushima Daiichi nuclear power plant) was analyzed, and the tendencies were discussed from the viewpoint of complexation. Moreover, chromatographic separation of U from the other elements was evaluated using the resin. The results indicate that this type of resin has high U separation selectivity. Consequently, the authors developed a novel strategy involving the use of this resin for U separation in conjunction with chlorination of hardly soluble materials by using atmospheric-pressure nonthermal plasma. This strategy is also proposed to be applicable for stabilizing waste forms and for nuclear fuel material accounting, such as for debris from the Fukushima nuclear power plant and other complex hardly soluble nuclear materials.     

Separation of Uranium from Fission Products

A simple procedure has been developed for separating U from F.P. and neutron-irradiated Th. The separation is performed with tri-n-butyl phosphate in a system of dodecane-mixture of sulfuric acid and aluminium nitrate.

Uranium dioxide was irradiated with 20 MeV bremsstrahlung, which produced both ²³⁷U and F.P. The target was dissolved in dilute nitric acid and U was extracted into the organic phase of the above mentioned system. Finally, U in the organic phase was back-extracted into an aqueous phase. The γ-ray spectrum and decay curve of the separated U fraction show no radioactive nuclides other than U isotopes and its decay products.

This method can also be applied to the preliminary separation of ²³⁸U from neutron- irradiated Th.

The distribution ratios (Kd) for U, Th and some other elements in the extraction system are also given.     

The separation of Zr95, Nb95, and Ru106 from a mixture of fission products by extraction with tributyl phosphate

We have studied methods for preparing radiochemically pure isotopes of Zr⁹⁵, Nb⁹⁵, and Ru¹⁰⁶ by a previously described [1] general scheme for the separation of fragmentary radioactive elements.We mainly consider regularities which were established in a study of the extraction of zirconium, niobium, and ruthenium by tributyl phosphate (TBP).Ruthenium is extracted by TBP after preliminary concentration on the sulfides of metals.Niobium and zirconium are separated by successive reextraction of niobium by hydrogen peroxide and zirconium by oxalic acid.Translated from Atomnaya Énergiya, Vol. 15, No. 1, pp. 23–30, July, 1963     

Production of weighable quantities of Tc99 from molybdenum irradiated with neutrons

Various methods of separating technetium from molybdenum anhydride irradiated with thermal neutrons in a nuclear reactor have been investigated. A method has been devised and tested In laboratory conditions for the concentration and separation of technetium, based on the co-precipitation of technetium with difficultly soluble phosphates and on Chromatographic purification. Milligram quantities of Tc⁹⁹ have been obtained by this method. The separated technetium has been identified by means of spectrum analysis, and the absolute activity and maximum energy of its ß -radiations have also been measured. Some chemical properties of technetium have been studied.Presented on March 9, 1957 at the All-Union Radiochemistry Conference in Leningrad.     

钚及其裂变产物钯 、银 、镉 、锡 、锑 、锆的系统分离方法

描述了钚及其6种裂变产物钯、银、镉、锡、锑、锆的系统分离方法：在强碱性阴离子交换树脂柱上将盐酸介质的辐照靶溶解液中的这些元素分为5组,然后再针对各组目标元素进行分离和纯化,可简便快速地从同一份靶溶解液中分离以上7种元素。采用辐照铀靶对分离方法进行了验证,结果表明,分离流程对6种裂变产物的化学回收率均大于70%,对γ谱仪测量干扰的主要核素去污因子均大于1.0×10³,可满足²³⁹Pu裂变谷区核素裂变产额测量对化学分离的要求。     

钚及其裂变产物钯、银、镉、锡、锑、锆的系统分离方法

描述了钚及其6种裂变产物钯、银、镉、锡、锑、锆的系统分离方法：在强碱性阴离子交换树脂柱上将盐酸介质的辐照靶溶解液中的这些元素分为5组，然后再针对各组目标元素进行分离和纯化，可简便快速地从同一份靶溶解液中分离以上7种元素。采用辐照铀靶对分离方法进行了验证，结果表明，分离流程对6种裂变产物的化学回收率均大于70%，对γ谱仪测量干扰的主要核素去污因子均大于1.0×10³，可满足²³⁹Pu裂变谷区核素裂变产额测量对化学分离的要求。     

Isotope composition of the rare earth elements formed in the fission of uranium,thorium, and bismuth nuclei by protons with energy of 680 Mev

The results are given of a radiochemical investigation, carried out in 1954, of the rare earth elements formed by the fission of uranium, thorium, and bismuth nuclei by 680 Mev protons, The main attention was devoted to methods of isolation and separation of these elements. In a search for the optimum conditions of separation by ion exchange chromatography, the effect of the nature of the complex former (ammonium acetate, nitrate, oxalate, and lactate), the pH of the eluant, and the concentration of the rare earth elements on the degree of separation was studied. A technique was evolved whereby it was possible to isolate the radioisotopes of almost all the rare earth elements and to determine the yields of some of them. In addition, the formation of a new samarium isotope with a half-life of about 20 days is presumed.     

Method for Calculating and Optimizing a Cascade for Separating a Mixture of Isotopes with Impurities

A method is proposed for calculating and optimizing a cascade separating a mixture of isotopes with impurities. The approach developed optimizes a cascade with arbitrary mixture composition and a large difference in the masses of the components being separated. The application of this method to the separation of a binary mixture of uranium isotopes with impurities is examined. The properties of cascades optimized with respect to the criterion of minimum total flux are investigated. Their correspondence to R cascades with symmetric steps is analyzed. A comparison is made with optimal cascades with no impurities. It is shown that multicomponent potentials can be used to estimate the separation efficiency for a binary mixture of isotopes with an impurity.__________Translated from Atomnaya Energiya, Vol. 98, No. 4, pp. 293–300, April, 2005.     

Nuclide separation from spent nuclear fuels by using tertiary pyridine resin

The newly nuclide separation system from spent nuclear fuels is proposed. The proposed separation system consists of recovery of nuclear fuel elements, separation of trivalent minor actinide from lanthanide, and separation of some fission products such as strontium. This separation system is based on the chromatographic technique using the tertiary pyridine resin. Separation experiments using mixed oxide fuel highly irradiated in fast reactor “Joyo” were carried out. The recovery of plutonium, the separation of minor actinide from fission products including lanthanides, and the separation of americium and curium were achieved. The recovery or removal of platinum group elements and technetium was also investigated, and the removal of these elements prior to the main reprocessing process has been proposed.     

Sequential Ion-Exchange Separation of Heavy Elements and Selected Fission Products for Burnup Measurement

A sequential ion-exchange separation method was developed for use in burnup measurements of nuclear fuels. Group separation by anion-exchange resin column with hydrochloric acid solutions containing small amounts of nitric acid and hydrochloric acid was followed by various cation and anion- exchange processes. The heavy elements, such as U, Np and Pu, and some fission products selected as burnup monitors, such as Cs, Mo and Nd, could be sequentially and quantitatively separated from a sample taken from spent fuel. The recovery of these elements through the separation processes were examined. The sampling ratio of an aliquot in reference to the whole fuel specimen was determined by adding as sampling monitor a known amount of Cu to the sample during dissolution. The validity of the ion-exchange separation technique for routine analysis for burnup measurements is also discussed.     

Separation of Plutonium-238 from Fission Products by Solvent Extraction Using HDEHP

A procedure for separating ²³⁸Pu from a Np sample irradiated with neutrons is described. Rapid separation of Pu by HDEHP solvent extraction was attempted, and without adjusting its valency states in the dissolver solution of the sample. Both Pu(IV) and Pu(VI) were extracted along with Np from the HNO₃ solutions of various concentrations. The Pu and Np extracted in the organic solution were back-extracted with oxalic acid solutions. The decontamination factors of the crude products were of the order of 10² for gross γ-activity. The Pu in the products was separated from Np by means of ion exchange resin columns. Approximately 0.5 mg of ²³⁸Pu was obtained with an efficiency exceeding 95%.     

Studies on Focusing Chromatography

U, Np and F.P. in neutron irradiated uranium oxide were separated from each other using focusing chromatography. Individual nuclides were determined hy γ-ray spectrometry. As complex forming agent, 1.0M lactic acid and 0.05M nitrilotriacetic acid (NTA) were tried, and various mixtures of hydrochloric acid and sodium chloride as complex decomposing agent.

The efficacy of the forming and decomposing agents for the separation was observed with various solutions of electrolytes. In all cases, Tc, Mo, Ru, Rh, I, Zr and one fraction of Np migrated toward the anode. As the hydrogen ion concentration of the complex decomposing solution was decreased, these nuclides became more easily separable. Te and a small amount of Np generally remained in the area of the spot where the sample had been introduced. Sr and Ba always migrated toward the cathode. Rare earth elements were sensitive not only to the kind of complex forming agent used but also to the pH value of the decomposing solution. Np manifested three different modes of focusing characteristic, their proportions varying with the conditions of the separation system.

The concluding result was that the system of 0.1M sodium chloride—1.0 M lactic acid was found the most suitable for separating these multi-component mixtures.     

Separation of Antimony-125 in Fission Products

Most of the known radioactive nuclides of antimony produced by neutron irradiation of uranium have fission yields below 1% and have half-lives below 60 days. An exception is ¹²⁵Sb with a half-life of 2.7 yr, which raise its relative importance among the fission products with lapse of time after irradiation, and after 1 yr of cooling, its radioactivity is no longer negligible. This circumstance has led to its being separated from such sources as fall-out. No studies have so far been reported on using the nitrate system for this separation, though it is utilized in the reprocessing of spent fuel and in the dissolution of uranium samples. The present work describes a method of separating ¹²⁵Sb from fission products with use made of silica gel—nitric acid system, and an example of its application to the separation of ¹²⁵Sb from the spent fuel of JPDR-1. The fuel was irradiated from Oct. 1963 to Sep. 1969. The amount of ¹²⁵Sb measured after separation was (1.7± O.19)×10^?1Ci/gU at June 1972.     

Optimum Cut of Separating Element as Function of Total Separation Factor and Mole Fraction of Component to be Separated

For a simplified model of separating elements where a total separation factor αβ is independent of values of cut ? (0≤?≤1), an optimum cut ?_opt in the sense that the cut makes a separative power δU maximum, was derived in terms of αβ, and a mole fraction x_F (0≤x _F≤1) of the component to be separated. When values of x^F is nearly equal to zero, the optimum cut ?^xF _opt?0 decreases and approaches to near 0, as the total separation factor becomes larger. On the contrary, when x_F is nearly equal to 1, the optimum cut ?^xF _opt?1 was found from calculation to be 1-?^xF _opt?0, and increases and approaches to near 1, as the total separation factor becomes larger. Moreover, in the case of x_F =0.5, the optimum cut is 0.5 regardless of αβ. Generally, the optimum cut ?_opt(x_F ) was solved to be in the form of a linear interpolation of the boundary values, ?x^F _opt?0 and ?x^F _opt?1.     

Separation of the stable isotopes of boron

This article describes methods for separating the stable isotopes of boron. Three of them can be used to prepare concentrates of B¹⁰ isotope in industrial quantities. The method of chemical exchange has a comparatively high coefficient of separation ( = 1.03); however, the production rate of the apparatus is small due to the high molecular weight of the complex. The preparation of B¹⁰ by fractionating BF₃ offers possibilities ( = 1.0075); however this process must be carried out at a temperature of –l00 °C, and requires the use of large quantities of liquid air. The method of separation by fractionation of BCl₃ has a low value of (1.003), but can be carried out at atmospheric pressure and room temperature.     

Rapid Separation of Fission-Product Rare Earths and Yttrium by Electromigration

The separation of fission-product La, Ce, Pr, Nd and Y was performed in 4 min by electro-migration. As complexing agent, nitrilotriacetic acid was used at two concentrations of the NTA ion selected so as to optimize mutual separation of the elements. Rapid location of the zone on the paper strip reached after migration was obtained by color reaction of the carrier added to the irradiated sample. The purity of the elements thus separated was checked by γ-ray spectrometry. At the higher concentration of NTA ([[NTA]=1.0×10^?2M, pH=4.1) under which all of rare earths and Y migrate toward the anode, the elements were well separated from each other, although appreciable amounts of other F.P. were also found in each fraction. On the other hand, at the lower NTA concentration ([NTA]=4.0 × 10^?3M, pH=2.0) where the F.P. all migrate toward the cathode, each zone contained the isotopes of the respective element almost exclusively, except for a few small photo-peaks attributable to other elements. In each fraction, the 92.5 min ¹⁴²La and 14 min ¹⁴³La, the 14 min ¹⁴⁶Ce (24 min ¹⁴⁶Pr), the 24 min ¹⁴⁶Pr and 12 min ¹⁴⁷Pr, the 1.8 hr ¹⁴⁹Nd, the 20.3 min ⁹⁴Y and 10.9 min ⁹⁵Y were identified from the energies of photo-peaks and their decay rates.     

AG-MP-1M在氢溴酸体系中分离镉的方法

Cd同位素质谱分析中存在的质谱干扰有¹¹⁰Pd、⁷⁰Zn⁴⁰Ar、⁹⁵Mo¹⁶O、¹¹²Sn、¹¹³In、¹¹⁴Sn和⁹⁸Mo¹⁶O等。因此在分析前应将Cd元素与产生干扰的其他元素分离,以避免测量过程中的质谱干扰。目前分离Cd的常用方法通常采用盐酸体系,分离过程中存在Sn与Cd分离困难的问题,操作步骤繁琐且淋洗体积大。针对上述问题,本工作开发了用阴离子树脂AG-MP-1M在氢溴酸体系中分离Cd的方法,采用混合酸消解的方法将样品消解完全后,用1 mL 0.25 mol/L的氢溴酸溶解样品并上柱,用3 mL 0.25 mol/L氢溴酸淋洗后可将绝大部分Sn、Zn、Mo等元素分离,然后用1 mL 2 mol/L盐酸和3 mL 0.5 mol/L盐酸淋洗Pb等其它元素,最后,用3 mL 0.002 mol/L盐酸解吸Cd。实验表明Cd回收率可达99.1%,且能将Sn、Zn、Mo等产生干扰的元素完全分开。采用此方法分离Cd总的淋洗体积为10 mL,和之前文献报道的方法相比大量减少了淋洗体积。该方法提高了Cd的分离效果,节约了实验流程时间,可用于Cd含量测试和Cd同位素测试的前处理工作中。     

反应堆中子活化-多元素亚化学计量分离法测定生物和半导体材料中的Cu,Na,Au

近年来,由于Ge(Li)γ谱仪的普遍使用,亚化学计量分离法朝着多元素组分离的方向发展,亦可同时测定多种元素。 Elek和Kukala等人作了多元素亚化学计量组分离的理论研究,提出了关于MA_n类型的金属螯合物的多元素亚化学计量分离的关系式。如果指定:定量萃取(≥99％)金属离子M_i,同时保证亚化学计量(50％)萃取金属离子M_s,而金属离子M_1不被萃取,留在水相(≥99％);当等体积萃取时,有下述关系式: