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.     

Current Progress in R&D of Fluoride Volatility Method

Fluoride Volatility Method is regarded to be a promising advanced pyrochemical reprocessing technology, which can be used for reprocessing mainly of oxide spent fuels coming from current LWRs or future GEN IV fast reactors. The technology should be chiefly suitable for the reprocessing of advanced oxide fuels with inert matrixes of very high burn-up and short cooling time, which can be hardly reprocessed by hydrometallurgical technologies. Fluoride Volatility Method is based on direct fluorination of powdered spent fuel with fluorine gas in a flame fluorination reactor, where the volatile fluorides (mostly UF6) are separated from the non-volatile ones (trivalent minor actinides and majority of fission products). The subsequent operations necessary for partitioning of volatile fluorides are condensation and evaporation of volatile fluorides, thermal decomposition of PuF6 and finally distillation and sorption used for the purification of uranium product.     

Poly(vinylidene fluoride) as a reagent for the synthesis of K2NiF4-related inorganic oxide fluorides

In this paper we report an improved route to the synthesis of K₂NiF₄-related inorganic oxide fluorides, such as Sr₂TiO₃F₂ and Ca₂CuO₂F₂ using low-temperature fluorination of precursor oxides with poly(vinylidene fluoride). Use of this fluorinating agent results in high quality samples, without SrF₂ or CaF₂ or other impurities, which are commonly seen for alternative fluorination routes.     

New approaches to processing and fabrication of oxide nuclear fuels

It was shown that, in contrast to the Purex process using aggressive and environmentally hazardous 8M HNO₃ solutions for dissolving spent oxide nuclear fuel (SNF), this fuel can be easily dissolved in aqueous subacid ([H⁺] ∼0.1 M) solutions of Fe(III) nitrate (chloride) with partial separation of uranium and plutonium from fission products (FP). The low acidity of the solutions obtained (pH ∼1) allows direct application of modern technologies of finishing processing of nuclear fuel by fluoride, carbonate, oxalate, or peroxide precipitation of uranium and plutonium. It was established that U(VI) is isolated from nearly neutral nitric acid solutions as a poorly soluble uranyl hydroxylaminate complex after adding hydroxylamine. It was shown that on thermal decomposition at 200–300°C under ambient atmosphere this compound converts into uranium dioxide. A similar approach was applied to obtain mixed oxide uranium-plutonium fuel (MOX fuel).     

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.     

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.     

Role of elemental fluorine in nuclear field

The preparation of fluorine gas by Henri Moissan by electrolysis of molten fluorides can be considered as one of the most important discoveries during the last centuries. Indeed, in addition to its use in many industrial fields (microelectronic, surface cleaning, pharmacology, medicine, …), fluorine gas is strongly involved in nuclear field for the preparation of UF₆. The latter allows the natural uranium enrichment via the gaseous diffusion process. Due to the increase of the energy demand in industrialised and emergent countries, the production of UF₆ and consequently of F₂ should increase drastically during the next decades. The aim of this paper is to summarise the evolution of the process to produce fluorine from its discovery to the present process. Few aspects on the researches done for a better understanding of the fluorine evolution reaction are presented. The use of fluorine in the nuclear field is also discussed.     

Recent developments in the preparation of high surface area metal fluorides

An overview of the main procedures for the preparation of fluorides with very high surface areas is given. Three processes are outlined: (i) plasma fluorination, (ii) sol–gel route and (iii) oxidative decomposition of inorganic precursors. From all three processes nanostructured metal fluorides with 100–400 m² g⁻¹ can be obtained. Prevention of the local overheating during fluorination seems to be the key factor to obtain the high surface area fluorides. TEM investigations of AlF₃ and CrF₃ obtained by oxidative decomposition revealed considerable differences in their morphologies and crystallinity. CrF₃ is completely amorphous and unstable under beam. AlF₃ contains an amorphous phase and nanocrystalline phases of α-AlF₃ and β-AlF₃. Nanocrystals are uniformly distributed within the amorphous phase. Also present are the rod-like nanostructures that consist of β-AlF₃ and are 5–10 nm wide.     

Novel method of synthesis of C60F48 with improved yield and selectivity

A novel two-stage method of preparation of C₆₀F₄₈ with 96% purity and 80% yield is reported. A C₆₀ embedded into a MnF₂ matrix is reacted with molecular fluorine under dynamic conditions, i.e. in flow of fluorine gas and with sublimation of volatile products, which results in formation of C₆₀F₃₄-C₆₀F₃₈ mixtures with >90% yield. Subsequent fluorination of the mixture thus obtained in the closed reactor at elevated pressure directly leads to the final product. C₆₀F₄₈ thus synthesized has been characterized by means of EI-MS, MALDI-MS, IR-spectroscopy and X-ray photoelectron spectroscopy (XPS). The problems of fullerene burning and degradation in the fluorine atmosphere are discussed.     

Studies on decontamination of spent ion exchange resin used for plutonium purification in PUREX stream

This paper deals with the studies on decontaminations of spent ion exchange resin used for purification of plutonium in PUREX process stream. Studies were carried out to optimize the chemical procedure for removal of plutonium and fission products activities form spent Ion Exchange resin. Different metal complexing reagents were tested for leaching out of radionuclides entrapped in irradiated spent ion exchange resin. The experimental results indicate that 0.01 M NaF solution was found the most suitable for removal of plutonium. The mixture of Na₂CO₃ and sodium salt of EDTA solution was found to be better for decontamination of spent ion exchange resin from beta and gamma activities. Optimized mixture of 0.5 M Na₂CO₃ and 0.1 M sodium salt of EDTA solution was found to be the most effective for fission product activities removal. After successive multiple contacts using these suitable reagents, the Pu and fission product activities in spent ion exchange resin were brought down to a minimum possible level, making it quite suitable for its long term storage.     

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 EURO-GANEX Process: Current Status of Flowsheet Development and Process Safety Studies

A new hydrometallurgical grouped actinide extraction process has been developed to separate the transuranic actinide ions from dissolved spent fuel solution (after an initial uranium extraction cycle). This “EURO-GANEX” process is aimed towards the homogeneous recycling of plutonium and minor actinides in a future closed fuel cycle. The separation process is based on the co-extraction of actinides and lanthanides from aqueous nitric acid into an organic phase followed by selective co-stripping of actinides. A suitable organic phase has been formulated and distribution ratios determined for lanthanides, actinides and some problematic fission products under extraction and stripping conditions. The process flowsheet has been proven on surrogate feed solutions as well as with spent fast reactor fuel; excellent recoveries of the actinides and good decontamination factors from the lanthanides and other fission products were obtained. A variation on the EURO-GANEX flowsheet (the “TRU-SANEX” process) has now been designed to produce separate Pu+Np and Am+Cm products for heterogeneous recycling. Progress on underpinning process chemistry and safety studies as well as flowsheet tests are summarized.     

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.     

Surface modification of carbon anodes for secondary lithium battery by fluorination

Recent results on the surface modification of petroleum cokes and their electrochemical properties as anodes of secondary lithium batteries are summarized. The surface of petroleum coke and those heat-treated at 1860-2800 °C were fluorinated by elemental fluorine (F₂), chlorine trifluoride (ClF₃) and nitrogen trifluoride (NF₃). No surface fluorine was found except only one sample when ClF₃ and NF₃ were used as fluorinating agents while surface region of petroleum coke was fluorinated when F₂ was used. Transmission electron microscopic (TEM) observation revealed that closed edge of graphitized petroleum coke was destroyed and opened by surface fluorination. Raman spectra showed that surface fluorination increased the surface disorder of petroleum cokes. Main effect of surface fluorination with F₂ is the increase in the first coulombic efficiencies of petroleum cokes graphitized at 2300-2800 °C by 12.1-18.2% at 60 mA/g and by 13.3-25.8% at 150 mA/g in 1 mol/dm³ LiClO₄-ethylene carbonate (EC)/diethyl carbonate (DEC) (1:1, v/v). On the other hand, main effect of the fluorination with ClF₃ and NF₃ is the increase in the first discharge capacities of graphitized petroleum cokes by ∼63 mAh/g (∼29.5%) at 150 mA/g in 1 mol/dm³ LiClO₄-EC/DEC.     

The determination of fluorine in rare earth fluorides by high temperature hydrolysis

The technique of pyrohydrolysis has been applied to the determination of fluorine in the fluorides of scandium, yttrium, and the lanthanons. These fluorides have been divided into two classes according to their rate of hydrolysis. Lutetium, ytterbium, cerium(III), scandium, gadolinium, terbium, dysprosium, holmium, erbium, and thulium florides can be hydrolyzed in 30 min or less. Yttrium, lanthanum, praseodymium, neodymium, samarium, and europium fluorides require from 45 to 150 min for complete hydrolysis. Accelerators, such as uranium oxide (U₃O₈), chromium(III) oxide, and a mixture of these oxides have been used successfully to reduce the tune required for quantitative hydrolysis of the fluorides in the latter group. The use of the correct accelerator reduces the hydrolysis time to 30 min or less for all these fluorides except lanthanum, praseodymium and neodymium.     

Partitioning of plutonium and uranium in aqueous medium using hydroxyurea as reducing agent

A new process for the partitioning of plutonium and uranium during the reprocessing of spent fuel discharged from fast reactor was optimised using hydroxyurea (HU) as a reductant. Stoichiometric ratio of HU required for the reduction of Pu(IV) was studied. The effect of concentration of uranium, plutonium and acidity on the distribution ratio (Kd) of Pu in the presence of HU was studied. The effect of HU in further purification of Pu such as solvent extraction and precipitation of plutonium as oxalate was also studied. The results of the study indicate that Pu and U can be separated from each other using HU as reductant.     

Low temperature preparation and uses of potent oxidizers

Because electronegativity of an oxidation state is low in an anion, salts of the high oxidation-state species [AgF₄]⁻ and [NiF₆]²⁻ can be easily made, at 0 °C, in liquid anhydrous HF (aHF) made basic with alkali fluorides. The containers are transparent fluorocarbon, and the F₂ is photo-dissociated. The [NiF₆]²⁻ salts, and the metastable binary fluorides NiF₄ and NiF₃, derived from them, are efficient fluorinating agents for the conversion of hydrido compounds to their fully fluorinated relatives. With F₂ in aHF made acidic with fluoride-ion acceptors (e.g. MF₅, M = As, Sb, Bi) attained oxidation-states are often lower (e.g. Ag^II, Au^II) because of the higher electronegativity in cations. Cationic Ag^III and Ni^IV species (derived from the anions) are sufficiently long-lived, and potent, to generate the most powerfully oxidizing hexafluorides of the second and third transition series (i.e. [MF₆], M = Pt, Ru, Rh). This synthesis is especially valuable for RhF₆, and has provided for the reinvestigation of the interaction of it with O₂. It is proposed that the unexpectedly large unit cell of O₂RhF₆ is a result of the presence of neutral O₂ and neutral RhF₆ as well as O₂⁺ and RhF₆⁻ in the lattice.     

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.     

UF6 plasma chemistry reconstitution experiments using high-temperature fluorine

Prior results indicate techniques have been developed for fluid mechanical confinement of high-temperature uranium hexafluoride (UF₆) plasma for long test times while simultaneously minimizing uranium compound deposition on the walls. Follow-on investigations were conducted to demonstrate a UF₆/argon injection, separation, and reconstitution system for use with rf-heated uranium plasma confinement experiments applicable to UF₆ plasma core reactors. A static fluorine batch-type regeneration test reactor and a flowing preheated fluorine/UF₆ regeneration system were developed for converting all the nonvolatile uranium compound exhaust products back to pure UF₆ using a single reactant. Pure fluorine preheat temperatures up to 1000 K resulted in on-line regeneration efficiencies up to about 90%; static batch-type experiments resulted in 100% regeneration efficiencies but required significantly longer residence times. A custom-built, ruggedized time-of-flight (T.O.F.) mass spectrometer, sampling, and data acquisition system permitted on-line quantitative measurements of the UF₆ concentrations down to 30 ppm at various sections of the exhaust system; this system proved operational after long-time exposure to corrosive UF₆ and other uranium halides.     

煤热解过程中含氮气相产物转化规律的实验研究

为了研究煤在热解过程中含氮气相产物的生成规律,在滴管炉反应系统中对四种原煤以及两种脱除矿物质煤样分别在500℃、700℃、900℃和1100℃进行了实验研究。结果表明,随着温度的升高,作为NO前驱物的HCN和NH3的收率随之增加,N2的收率也增加。煤种对含氮气相产物的生成规律也有着较大的影响,煤化程度比较低的煤在热解过程中,燃料氮向气相含氮产物的转化率较高;煤化程度比较高的煤转化率则偏低,大部分的氮缩聚在多环芳香结构中,成为焦炭氮。煤中的矿物质对燃料氮向N2的转化起到了促进作用,而对燃料氮向HCN和NH3的转化起到了抑制作用。