Organic solvent extraction of uranium from alkaline nuclear waste

There are proliferation issues with the Plutonium Uranium Redox Extraction process due to the possibility of recovering plutonium. The objective of this research was to evaluate different organic extraction ligands that can remove uranium from the nuclear waste and to determine the most effective organic solvent for extracting uranium only, from alkaline media. The results indicate that Alamine 336 in xylene has zero (0%) extraction capability for surrogate fission products at an optimum extraction time of 15 min. Aliquat 336 in xylene has an extraction percentage of 72% for uranium in 60 min. However, Aliquat 336 in toluene extracted 82% of the uranium from the feed solution after 30 min, decreasing to 76% after 60 min.

     

Feasibility of the recovery of uranium from alkaline waste by amidoximated grafted polypropylene polymer matrix

The amidoximated grafted polypropylene polymer matrix was prepared by post irradiation grafting of acrylonitrile (AN) onto thermally bonded non-woven matrix of poly(propylene) sheet using electron beams. This precursor polymer was reacted with hydroxylamine to convert AN to poly(acrylamidoxime) (AO) groups, and conditioned by treating them with 2.5 % KOH at 80 °C for 1 h. The polymer matrix was having the degree of AN grafting ~106 wt% and its subsequent conversion to AO groups ~70 %. The water uptake capacity of AO polymer matrix were found to be 100 ± 5 % (w/w). Quantitative recovery of uranium from alkaline waste (ammonium diuranate supernatant) solution was achieved by this polymer matrix. The other radionuclides present in the waste solution were not extracted by the polymer matrix. For all other radionuclides, the uptake was found to be <6 %.     

New reprocessing system for spent nuclear reactor fuel using fluoride volatility method

Our proposed spent nuclear fuel reprocessing technology named FLUOREX, which is a hybrid system using fluoride volatility and solvent extraction, meets the requirements of the future thermal/fast breeder reactors (coexistence) cycle. We have been done semi-engineering and engineering scale experiments on the fluorination of uranium, purification of UF₆, pyrohydrolysis of fluorination residues, and dissolution of pyrohydrolysis samples in order to examine technical and engineering feasibilities for implementing FLUOREX. We found that uranium in spent fuels can be selectively volatilized by fluorination in the flame type reactor, and the amount of uranium volatilized is adjusted from 90% to 98% by changing the amount of F₂ supplied to the reactor. The volatilized uranium is purified using UO₂F₂ adsorber for plutonium and purification methods such as condensation and chemical traps for fission products provide a decontamination factor of over 10⁷. Most of the fluorination residues that consist of non-volatile fluorides of uranium, plutonium, and fission products are converted to oxides by pyrohydrolysis at 600-800 °C. Although some fluorides of fission products such as alkaline earth metals and lanthanides are not converted completely and fluorine is discharged into the solution, oxides of U and Pu obtained by pyrohydrolysis are dissolved into nitric acid solution because of the low solubility of lanthanide fluorides. These results support our opinion that FLUOREX has great possibilities for being a part of the future spent nuclear fuel cycle system.     

The use of hydrochloric acid in separating95Zr−95Nb activity from uranium fission products on silica gel

Dissolution of a neutron-irradiated uranium target in a medium of 6N HCl containing a few drops of very dilute HNO₃ yielded a matrix solution which on running on a silica gel column allowed the complete adsorption of the⁹⁵Zr−⁹⁵Nb activity formed in the fission process. The⁹⁵Zr−⁹⁵Nb activity is cleanly and totally eluted with 0.5% oxalic acid solution. None of the uranium or the activity of the other fission products was found to be adsorbed on the column.     

Separation of gram quantities of uranium from fission products by extraction chromatography

The separation of gram quantities of uranium from fission products has been investigated by extraction chromatography. The separation which is based on the difference in distribution coefficients between uranium and the fission products on a tributyl phosphate (TBP) resin in nitric acid medium, was carried out by means of high acidity feed and stepwise elution on a TBP chromatography column. The results show that this technique is capable to separate 5 g of uranium from a large quantity of fission products. The recovery of uranium is more than 99%. The decontamination factors of g- and b-activities were 2.1^.10³ and 2.3^.10³, respectively.     

Chromatographic isolation of144Ce and144Pr from the wastes of irradiated uranium treatment

A two-step chromatographic technique was elaborated to isolate¹⁴⁴Ce,¹⁴⁴Pr from a solution of uranium fission products in 6M HNO₃. The oxidation to Ce(III) by bromate and selective adsorption of¹⁴⁴Ce(IV) on anion exchange column were used to concentrate and purify¹⁴⁴Ce. Some impurities of uranium,⁹⁵Zr,⁹⁵Nb,¹⁰⁶Ru remain in¹⁴⁴Ce solution after the first step of its isolation. The final purification is achieved by passing the 6M HNO₃ solution of¹⁴⁴Ce(IV) through the HDEHP-coated teflon column. The decontamination factors of¹⁴⁴Ce from main fission products are given. 7.2 mCi of (¹⁴⁴Ce+¹⁴⁴Pr) are recovered from each gram of irradiated uranium trioxide with the yield greater than 99%. An improvement of known generator was carried out to elute a purer¹⁴⁴Pr from maternal¹⁴⁴Ce(IV) adsorbed on the anion exchange column.     

Radiochemical determination of fission products of uranium

An analysis has been elaborated to determine the long-living γ-emitting fission products of uranium. It consists of a sodium bisulphate melt of the fission product solution or the U-fuel, followed by liquid-liquid extractions. Afterwards the isotopes are absolutely counted with a standardized 3″×3″ NaI crystal. The total γ-spectrum of the original fission product solution, taken with a NaI crystal or a Ge−Li detector, is also analyzed mathematically by mixed γ-spectrometry. From a short post-irradiation of the fission product solution the concentrations of both²³⁵U and²³⁸U are determined. The absolute amount of fission products related to the U-concentration allows the calculation of the percent atomic burn-up, the irradiation time, the cooling period, the flux of the reactor and the original degree of enrichment of the uranium. Research associate of the I. I. K. W.     

Radiochemical separation of99Mo from natural uranium irradiated with thermal neutrons

⁹⁹Mo was separated from uranium and insoluble fission product hydroxides. More than 98% of⁹⁹Mo radioactivity was extracted with bis (2-ethylhexyl)phosphoric acid. The organic phase was washed and⁹⁹Mo was back-extracted from the organic phase with NH₄OH solution. The percent recovery from the organic phase was 91% and the purity of⁹⁹Mo was more than 99%. Pure^99mTc was also extracted from the organic phase with a saline solution. Reversed-phase partition chromatography was used for the purification of⁹⁹Mo from¹³¹I and other fission products (10% HDEHP on kieselguhr bed).¹³¹I and other isotopes were quantitatively eluted with 0.1M H₂SO₄,⁹⁹Mo was eluted using a mixture of 0.5 M HCl and 30% H₂O₂.     

Determination of uranium fission interference factors for INAA

Instrumental neutron activation analysis (INAA) is a very suitable technique for the determination of several elements in different kinds of matrices. However, when the sample contains high uranium concentration this method presents interference problems of uranium fission products. The same radioisotopes used in INAA are formed in uranium fission. Among these radioisotopes are ¹⁴¹Ce, ¹⁴³Ce, ¹⁴⁰La, ⁹⁹Mo, ¹⁴⁷Nd, ¹⁵³Sm and ⁹⁵Zr. The purpose of this study was to evaluate uranium fission interference factors to be used in the INAA of environmental and geological samples containing high levels of U. The obtained interference factors agreed with literature reported values. The results point to the viability of using these experimentally determined interference factors for the correction of uranium fission products.     

Determination of the fast fission yield of141Pr using neutron activation analysis

A method is described for the determination of the fission yield of¹⁴¹Pr. This method was developed to determine the fast fission yield of¹⁴¹Pr in the Mark III loading (enriched uranium with about 2% zirconium) of the fast fission breeder reactor, EBR-1. The burnup of the fuel sample was determined using the previously reported fission yield of¹³⁷Cs. Praseodymium was separated from uranium, plutonium and other fission products by a combination of precipitation and ion exchange stages. Thereafter,⁵⁵Mn was added to serve as an internal flux monitor and praseodymium determined by neutron activation analysis. A precision of ±2% was obtained. Presented at the 15th Annual Meeting of the American Chemical Society, Miami Beach, Florida (USA), April 1967.     

A rapid radiochemical procedure for indium

A rapid radiochcmical procedure was developed for the separation of indium radionuclides from a mixed fission-product solution. An alcoholic pyridine solution is added to a uranium solution containing indium and tin carriers. The resulting tin precipitate is separated from the indium-containing solution by filtering through a cellulose membrane filter. The decontamination factor for tin is 2·10³. Other fission products are only partially removed. The chemical yield of indium is about 44%, and the time required for the separation is about 10 sec. After the tin-separated indium has decayed, the tin daughters of indium are removed from all the other fission products at a specified time and measured, so that the amount of indium present at the time of the tin precipitation is determined.     

Isolation of radiocesium from a fission product mixture

A method of¹³⁷Cs isolation from strongly, acidic solutions of fission products is described, in which vanadyl ferrocyanide is used as a selective ion exchanger for cesium. The effects of the acidity of medium and the carrier concentration on the quantitative yield of separation have been studied and convenient conditions have been found for¹³⁷Cs isolation from the solution of fission products formed after irradiating uranium with neutrons.     

Determination de l'enrichissement de l'uranium par comptage du rayonnement Cerenkov

A method for the measurement of²³⁵U in dilute uranium solutions based on Cerenkov radiation is described. It is applicable to solutions treated in fuel element production where the enrichment factor of the uranium solution is to be known, thus to solutions of uranyl nitrate not containing other fission products.      

低浓缩铀靶辐照后溶液中铀的化学种态及主要裂变元素的影响

利用化学种态分析软件CHEMSPEC计算了低浓缩铀靶辐照后溶液中铀(U)的化学种态分布及其主要裂变元素对U化学种态的影响。结果表明,在单组分体系中,pH值和铀酰浓度都会显著影响U的化学种态分布。随着铀酰浓度的增大,溶液中将会生成多核配合物;在较高的NO₃^-浓度下,U在溶液中主要以UO₂²⁺和UO₂NO³⁺的形式存在。CO₂对不同浓度铀的种态分布影响结果表明,当铀酰浓度较低时,铀的化学种态多以碳酸铀酰的形式存在;当铀酰浓度较高时,铀的化学种态多以氢氧铀酰或柱铀矿沉淀的形式存在。计算发现,当裂片元素Tc、I、Mo的浓度小于0.01mol·L^-1并分别以TcO₄^-、I^-、MoO₄^2-的种态存在时,这些裂片元素不改变铀的各化学种态的分布。     

低浓缩铀靶辐照后溶液中铀的化学种态及主要裂变元素的影响

利用化学种态分析软件CHEMSPEC计算了低浓缩铀靶辐照后溶液中铀(U)的化学种态分布及其主要裂变元素对U化学种态的影响。结果表明,在单组分体系中,pH值和铀酰浓度都会显著影响U的化学种态分布。随着铀酰浓度的增大,溶液中将会生成多核配合物;在较高的NO₃^-浓度下,U在溶液中主要以UO₂²⁺和UO₂NO₃⁺的形式存在。CO₂对不同浓度铀的种态分布影响结果表明,当铀酰浓度较低时,铀的化学种态多以碳酸铀酰的形式存在;当铀酰浓度较高时,铀的化学种态多以氢氧铀酰或柱铀矿沉淀的形式存在。计算发现,当裂片元素Tc、I、Mo的浓度小于0.01 mol·L^-1并分别以TcO₄^-、I^-、MoO₄^2-的种态存在时,这些裂片元素不改变铀的各化学种态的分布。     

Synthesis and Characterization of Chitosan‐graft‐Poly(3‐(trimethoxysilyl)propyl methacrylate) Initiated by Ceric (IV) Ion

The grafting of 3‐(trimethoxysilyl)propyl methacrylate (TMSPM) onto chitosan by ceric ion initiation was studied under homogeneous conditions in 2% acetic acid solution. The grafted polymer was characterized by FT‐IR, ¹H‐NMR, TGA and XRD and swelling studies. TGA results showed that the incorporation of TMSPM to the chitosan chains decreased the thermal stability of the grafted chitosan. Due to the grafting of TMSPM, the crystallinity of chitosan derivatives was found to be destroyed. The solubility of the grafted chitosan in water was improved. The effects of reaction conditions such as initiator concentration, monomer concentration, reaction temperature and reaction time were studied by determining the grafting parameters such as grafting and grafting efficiency. Under optimum conditions, the grafting parameters were achieved as 1440 and 97%, respectively.     

The determination of fission products in irradiated uranium by electrophoretic focussing of ions

Electrophoretic focussing of ions was applied to the separation of fission products present in solutions of nuclear uranium fuel irradiated in various European reactors. By combining two separation methods, all the long-lived fission products could be determined individually and quantitatively by counting with a NaI(T1) and a GM detector of known detection efficiency. Radiography and autoradiography were used for semi-quantitative purposes. The concentrations of²³⁵U and²³⁸U were determined from a short post-irradiation of the fuel solution and counting of¹⁴⁰Ba−¹⁴⁰La and²³⁹Np, respectively. An iterative calculus method is presented which allows calculation of the irradiation history of the fuel solution from the above analyses. without any a priori knowledge.     

Comparative study of cesium adsorption on dioctahedral and trioctahedral smectites

Bentonites which are characterized by good rheological, mineralogical and chemical stability is considered used as sealing barriers in multibarrier Slovak system of deep geological repository for high-level radioactive waste and spent nuclear fuel. In Slovak Republic there are several significant deposits of bentonite, which are characterized by appropriate adsorption properties and meet the geotechnical requirements for this type of barriers. Study of adsorption properties of bentonites and other smectites is an essential step for developing the migration model long-lived corrosion and activation products, and fission products of uranium. Nuclear wastes contain the most important nuclear fission products, radioisotopes ¹³⁴Cs and ¹³⁷Cs. The present paper investigates and compares the cesium adsorption properties of Slovak and North America bentonites composed mainly of dioctahedral smectite montmorillonite (J, L, SAz-1 and STx-1) and trioctahedral smectites saponite (SapCa-2) and hectorite (SHCa-1).     

Application of neutron activation analysis in a fission molybdenum separation study

The separation of⁹⁹Mo from low-enriched uranium (LEU, 19.5%²³⁵U) targets was evaluated using natural uranium (NU) and non-radioactive tracers. Neutron activation analysis was used to determine (1) the efficiency of molybdenum recovery and (2) the decontamination factor of numerous fission product elements from the molybdenum product. Using NU and non-radioactive elements simplified procedures and allowed tests to be completed in a fume hood instead of a shielded cell. During activation of the non-radioactive tracers, uranium fission occurs, which can interfere with subsequent gamma-ray analysis. A comparison was made of the interferences caused by these fission products from both NU and LEU.     

Fission track–secondary ion mass spectrometry as a tool for detecting the isotopic signature of individual uranium containing particles

A fission track technique was used as a sample preparation method for subsequent isotope abundance ratio analysis of individual uranium containing particles with secondary ion mass spectrometry (SIMS) to measure the particles with higher enriched uranium efficiently. A polycarbonate film containing particles was irradiated with thermal neutrons and etched with 6 M NaOH solution. Each uranium containing particle was then identified by observing fission tracks created and a portion of the film having a uranium containing particle was cut out and put onto a glassy carbon planchet. The polycarbonate film, which gave the increases of background signals on the uranium mass region in SIMS analysis, was removed by plasma ashing with 200 W for 20 min. In the analysis of swipe samples having particles containing natural (NBL CRM 950a) or low enriched uranium (NBL CRM U100) with the fission track–SIMS method, uranium isotope abundance ratios were successfully determined. This method was then applied to the analysis of a real inspection swipe sample taken at a nuclear facility. As a consequence, the range of ²³⁵U/²³⁸U isotope abundance ratio between 0.0276 and 0.0438 was obtained, which was higher than that measured by SIMS without using a fission track technique (0.0225 and 0.0341). This indicates that the fission track–SIMS method is a powerful tool to identify the particle with higher enriched uranium in environmental samples efficiently.