Determination of oxygen to uranium ratio in irradiated uranium dioxide by controlled potential coulometry

This work reports the determination of oxygen to uranium (O/U) ratio in irradiated UO_2+x fuel pellet of burnup of ca. 34 GWd/t by controlled potential coulometry. The method is based on the dissolution of the nuclear fuel in strong phosphoric acid (SPA) at 180–190 °C under an inert atmosphere. After dissolution, 8% sulphuric acid is added in order to obtain a 20% SPA in 8% sulphuric acid. A controlled potential coulometric determination of uranium(VI) is carried out at ?0.60V vs. ferri-ferrocyanide. The uranium(IV) contained in an aliquot of the fuel solution is oxidised to uranium(VI) with cerium(IV) sulphate, and the total uranium content is then determined by coulometry. Optimum experimental conditions have been established using simulated irradiated fuel solution containing various fission products which include cerium, tellurium, palladium, ruthenium, molybdenum and zirconium. Interference of the fission products and the possible removal of their interferences by preelectrolysis at +0.5 V vs. saturated calomel electrode (SCE) have been investigated. The accuracy of the coulometric method is confimed by polarographic measurement using several unirradiated UO_2+x fuel of known stoichiometry.     

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.     

Direct burn-up determination of fast reactor mixed oxide (MOX) fuel by preferential evaporation of interfering elements

Spent fuel of uranium–plutonium mixed oxide (MOX) from sodium cooled Fast Breeder Test Reactor (FBTR) was analyzed for at.% burn-up by preferential evaporation method. A sequential pattern of analysis of fission monitor Nd and heavy elements, U and Pu provided an un-interfered isotopic composition. Concentrations of individual elements were determined by isotopic dilution mass spectrometry. The proposed method provides at.% burn-up with an uncertainty of about 4% (compared to ASTM method), is less time consuming, does not involve any chemical separation, reduction in radioactive waste and substantial reduction in the radiation exposure to analyst.     

Simultaneous Plutonium and Uranium Isotopic Analysis from a Single Resin Bead - A Simplified Chemical Technique for Assaying Spent Reactor Fuels

Abstract

The method uses basic anion resin to adsorb plutonium and uranium from 7–8 M HNO₃ solutions containing dissolved spent reactor fuels. After equilibrating the resin with the solution, a single bead is used to determine the isotopic composition of plutonium and uranium on sample sizes as small as 10^?9 to 10^?10 g of each element per bead. Isotopic measurements are essentially free of isobaric interferences and fission product contamination in the mass spectrometer is eliminated. A very small aliquot of dissolver solution containing 10^?6 g of U and 10^?8 g of Pu is sufficient sample for chemically preparing several resin beads. A single prepared bead is loaded onto a rhenium filament and analyzed in a two-stage mass spectrometer using pulse counting for ion detection to obtain the high sensitivity required. Total quantity of the elements, in addition to isotopic abundances, can be determined by isotope dilution. Other areas where the method may be useful are: in plutonium production, isotope separations, and for trace detection of contamination on reactor parts.     

New approaches to reprocessing of oxide nuclear fuel

Dissolution of UO₂, U₃O₈, and solid solutions of actinides in UO₂ in subacid aqueous solutions (pH 0.9–1.4) of Fe(III) nitrate was studied. Complete dissolution of the oxides is attained at a molar ratio of ferric nitrate to uranium of 1.6. During this process actinides pass into the solution in the form of U(VI), Np(V), Pu(III), and Am(III). In the solutions obtained U(VI) is stable both at room temperature and at elevated temperatures (60 °C), and at high U concentrations (up to 300 mg mL^?1). Behavior of fission products corresponding to spent nuclear fuel of a WWER-1000 reactor in the process of dissolution the simulated spent nuclear fuel in ferric nitrate solutions was studied. Cs, Sr, Ba, Y, La, and Ce together with U pass quantitatively from the fuel into the solution, whereas Mo, Tc, and Ru remain in the resulting insoluble precipitate of basic Fe salt and do not pass into the solution. Nd, Zr, and Pd pass into the solution by approximately 50 %. The recovery of U or jointly U + Pu from the dissolution solution of the oxide nuclear fuel is performed by precipitation of their peroxides, which allows efficient separation of actinides from residues of fission products and iron.     

A universal solvent extraction-titration method for the rapid and accurate determination of uranium in complex solutions

A rapid and accurate method has been developed for the determination of uranium in complex solutions produced in the recovery of uranium from nuclear fuels. These solutions usually contain 0.1–20 g U l^-1 and high concentrations of aluminum nitrate in addition to a variety of cations and anions associated with fuel-element constituents and dissolution media. The method involves the solvent extraction of uranium from an acid-deficient aluminum nitrate-tetrapropylammonium nitrate solution into 2% tributy1 phosphate in n-amy1 acetate. The uranium is then backextracted with a concentrated phosphoric acid solution, and titrated by the method of Davies and Gray. The uranium extraction efficiency for sample solutions weighing up to 50 g is 99.9% or better, and the limit of error per analysis with 95% confidence is ±0.6%. No prior sample preparation is necessary, no expensive equipment is required, and even unskilled personnel can do duplicate analyses in 1.5 h.     

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.      

Fluoride volatility method for reprocessing of LWR and FR fuels

Fluoride volatility method is based on direct fluorination of powdered spent fuel with fluorine gas in a flame fluorination reactor, where the volatile fluorides (represented mainly by UF₆, partially NpF₆) are separated from the non-volatile ones (e.g. PuF₄, AmF₃, CmF₃, fluorides of majority of fission products), the objective being to separate a maximum fraction of uranium component from plutonium, minor actinides and fission products. The current research and development work in the area of fluoride volatility method is focused on the experimental program carried out at the semi-technological line called FERDA, which is a follow-up of the previous FREGAT-2 technology. The experimental test program, launched in 2004 by the Nuclear Research Institute ?e? plc, has been focused mainly to the study of flame fluorination process, which is considered to be the crucial unit operation of the technology. The fluorination experiments were realized in the first instance with pure uranium oxide fuel and later on with simulated spent oxide fuel. Follow-on tests are planed with oxide fuels with inert matrixes. The experimental program is further supplemented by the system studies focused mainly to the process flow-sheet design and calculations and to the requisite modification of some apparatuses for the future verification of the process with irradiated fuel in hot conditions.     

Burn-up measurements on nuclear reactor fuels using high performance liquid chromatography

Burn-up measurements on thermal as well as fast reactor fuels were carried out using high performance liquid chromatography (HPLC). A column chromatographic technique using di-(2-ethylhexyl) phosphoric acid (HDEHP) coated column was employed for the isolation of lanthanides from uranium, plutonium and other fission products. Ion-pair HPLC was used for the separation of individual lanthanides. The atom percent fissions were calculated from the concentrations of the lanthanide (neodymium in the case of thermal reactor and lanthanum for the fast reactor fuels) and from uranium and plutonium contents of the dissolver solutions. The HPLC method was also used for determining the fractional fissions from uranium and plutonium for the thermal reactor fuel.     

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.     

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。     

Neptunium estimation in the spent fuel dissolver solution by inductively coupled plasma-atomic emission spectroscopy

This paper deals with the optimization of experimental conditions for the estimation of Np in spent fuel dissolver solution using 2-thenoyltrifluoroacetone (HTTA) as extractant. The quantitative extraction of Np from the dissolver solution employing 0.5 M HTTA/xylene was followed by its estimation by Inductively Coupled Plasma-Atomic Emission Spectroscopy (ICP-AES) after stripping it from the organic phase with 8 M HNO₃. The reliability of the method was checked by standard addition technique. The method is precise and accurate yielding Np analytical recovery of 99 ± 1%.     

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.     

Sorption of uranium from alkaline waste onto radiation grafted phosphoryl-g-Teflon matrix

High energy ⁶⁰Co γ-radiation was used to graft glycidylmethacrylate onto Teflon scrap through mutual radiation grafting technique. The epoxy ring of grafted polyGMA chains were later converted to U selective phosphoryl group, chemically. The grafted matrix was used as solid–liquid adsorbent of uranium from alkaline waste solution. More than 98% recovery of uranium from alkaline waste (~pH 8) solution was achieved. The effect of grafting extent on adsorption kinetics was also investigated. The selectivity of uranium extraction over other fission products was established. The uptake of other fission products was <5% for equilibration time of ~1 h.

     

Recovery and purification of plutonium in lean plutonium product stream obtained during reprocessing of fast reactor spent fuels

A method has been developed for final purification of plutonium from uranium and fission products of high beta gamma activity. This method involves selection of a suitable ion exchange resin for the purification of plutonium in order to deliver a quality PuO₂ product. The effect of the concentration of uranium and plutonium, effect of increased loading of uranium and number of bed volumes for effective washing, which are some of the parameters that generally affect the recovery and purification of plutonium were investigated. An excellent decontamination factor for fission products has been achieved by this anion exchange process which in turn delivered an excellent PuO₂ product quality in terms of purity and associated beta gamma activity with low personnel radiation exposure.     

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.     

A fast and sensitive method for the determination of uranium in the Thorex Process streams

This paper presents a simple, rapid and sensitive radiometric method for the determination of uranium in Thorex Process stream containing large amount of thorium. This method involves the extraction of uranium into 0.05M tri-n-octyl phosphine oxide (TOPO) in xylene at 2M HNO₃. The extraction of thorium is prevented by masking them with suitable quantity of fluoride ions. The optimum experimental parameters for extraction of ²³³U were evaluated and using the most suitable experimental conditions the extracted uranium is radiometrically determined by α-counting in proportional counter with a prior knowledge of specific activity of uranium. Simultaneously in the same sample uranium was determined by spectrophotometric method using 2-(5bromo-2 pyridylazo)-5-diethylaminophenol (Bromo-PADAP) as chromogenic reagents. Simulated as well as actual samples of dissolver, conditioner and raffinate tank of Thorex stream have been analyzed by both these methods. The method was tested for as low as 0.15 μg of uranium and the results of these analyses were found to be satisfactory within the experimental limits.     

Determination of uranium,plutonium and selected fission products in the coolant of the KS-150 reactor

The method of monitoring of U, Pu and some fission products (^103,106Ru,^134,137Cs and^141,144Ce) in gaseous CO₂ coolant is described. The method is based on the retention of the radionuclides studied by membrane filters built in by-pass of the burst-cartridge detection (BCD) system. The purpose of the present study was the determination of U, Pu in CO₂ and the verification of the possibility of the indirect monitoring of U and Pu contents in the coolant, using the gamma-spectrometric determination of selected fission products retained by the filter. For calibration of the proposed method after decomposition of the filters, uranium was determined spectrophotometrically using Arsenazo III, plutonium was determined radiometrically after its separation by extraction with 2-thenoyltrifluoracetone and the fission products were determined by gamma-spectrometry. From the results obtained it follows that a correlation exists between the U and Pu content in the coolant and the activity of certain fission products retained on the filter.     

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.