Pyrochemical Reprocessing Research Articles

Electrochemical Properties of Uranyl Ion in Ionic Liquids as Media for Pyrochemical Reprocessing

We have proposed a new reprocessing process by using ionic liquids (ILs) instead of molten salts of alkali chlorides in pyrochemical process. In the proposed process, spent nuclear fuels are dissolved in ILs by using Cl2 as an oxidant, and UO2 2+ and PuO2 2+ ions in ILs are recovered as UO2 and PuO2 by electrochemical reduction. In order to examine applicability of ILs as media for reprocessing, we have studied electrochemical behavior of UO2 2+ in 1-butyl-3-methylimidazolium chloride (BMICl), 1-butyl-3-methylimidazolium tetrafluoroborate (BMIBF4), and 1-butyl-3-methylimidazolium nonafluorobutanesulfonate (BMINfO). Electrochemical properties of uranyl chloride dissolved into ILs were examined by cyclic voltammetry. In BMICl, an almost reversible redox couple was observed, and the formal potential and the diffusion coefficient were evaluated as _0:758V vs. Ag/AgCl and 4:8 × 10−8 cm2s−1, respectively. On the other hand, the electrochemical reactions of UO2 2+ in BMIBF4 and BMINfO were irreversible. In BMINfO, some reduction peaks and one sharp oxidation peak were observed in the range of −0:6∼–0:2V and around 0.85V vs. Ag/AgCl, respectively. The reduction and oxidation peaks were assigned to multi step reduction of UO2 2+ to U(IV) via U(V) and/or direct reduction of UO2 2+ to U(IV), and the oxidative dissolution of the resulting U(IV) compounds, respectively. The electrochemical reduction of UO2 2+ in BMINfO at −1:0V vs. Ag/AgCl produced the deposits on a carbon electrode as a cathode. Analyses of the deposits with the scanning electron microscope and the energy dispersive X-ray spectrometer indicated that the deposits are compounds containing uranium, oxygen, and chlorine. As a result, it is expected that the UO2 2+ in IL can be recovered electrolytically as uranium compounds such as UO2 and uranium oxychlorides.

A glass-encapsulated calcium phosphate wasteform for the immobilization of actinide-, fluoride-, and chloride-containing radioactive wastes from the pyrochemical reprocessing of plutonium metal

Chloride-containing radioactive wastes are generated during the pyrochemical reprocessing of Pu metal. Immobilization of these wastes in borosilicate glass or Synroc-type ceramics is not feasible due to the very low solubility of chlorides in these hosts. Alternative candidates have therefore been sought including phosphate-based glasses, crystalline ceramics and hybrid glass/ceramic systems. These studies have shown that high losses of chloride or evolution of chlorine gas from the melt make vitrification an unacceptable solution unless suitable off-gas treatment facilities capable of dealing with these corrosive by-products are available. On the other hand, both sodium aluminosilicate and calcium phosphate ceramics are capable of retaining chloride in stable mineral phases, which include sodalite, Na 8(AlSiO 4) 6Cl 2, chlorapatite, Ca 5(PO 4) 3Cl, and spodiosite, Ca 2(PO 4)Cl. The immobilization process developed in this study involves a solid state process in which waste and precursor powders are mixed and reacted in air at temperatures in the range 700–800 °C. The ceramic products are non-hygroscopic free-flowing powders that only require encapsulation in a relatively low melting temperature phosphate-based glass to produce a monolithic wasteform suitable for storage and ultimate disposal.

Retention studies on uranium, thorium and lanthanides with amide modified reverse phase support and its applications

The retention behaviour of uranium and thorium was investigated on modified reverse phase supports using 3-oxo-pentanedioicacid bis-[bis-(2-ethyl-hexyl)-amide (OPAEHA), 3-oxo-pentanedioicacid bis diisobutyl amide (OPAIBA) and bis-2-ethylhexyl succinamic acid (BEHSA). α-Hydroxy isobutyric acid (α-HIBA) was employed as the complexing reagent for elution. Elution profiles of uranium and thorium were studied as a function of the modifier concentration, mobile phase composition and its pH. Based on these investigations, a novel high performance liquid chromatography (HPLC) based separation technique was developed using BEHSA modified support for the isolation and quantitative determination of lanthanides as a group in uranium matrix. Hundreds of samples obtained from pyrochemical reprocessing of molten salts containing lanthanides in uranium matrix (e.g. 1:20,000) were separated and determined within 7 min using the coated support. The advantage of the present HPLC technique lies in the simultaneous separation and assay of total lanthanides and uranium whereas other analytical methods necessitate the separation of uranium matrix prior to lanthanide assay.

Electrochemical Properties of Uranyl Ion in Ionic Liquids as Media for Pyrochemical Reprocessing

Electrochemical Properties of Uranyl Ion in Ionic Liquids as Media for Pyrochemical Reprocessing

The evaluation of the pyrochemistry for the treatment of Gen IV nuclear fuels – Inert matrix chlorination studies in the gas phase or molten chloride salts

The structure of the fuels for the future Gen IV nuclear reactors will be totally different from those of PWR, especially for the GFR concept including a closed cycle. In these reactors, fissile materials (carbides or nitrides of actinides) should be surrounded by an inert matrix. In order to build a reprocessing process scheme, the behavior of the potential inert matrices (silicon carbide, titanium nitride, and zirconium carbide and nitride) was studied by hydro- and pyrometallurgy. This paper deals with the chlorination results at high temperature by pyrometallurgy. For the first time, the reactivity of the matrix towards chlorine gas was assessed in the gas phase. TiN, ZrN and ZrC are very reactive from 400°C whereas it is necessary to be over 900°C for SiC to be as fast. In molten chloride melts, the bubbling of chlorine gas is less efficient than in gas phase but it is possible to attack the matrices. Electrochemical methods were also used to dissolve the refractory materials, leading to promising results with TiN, ZrN and ZrC. The massive SiC samples used were not conductive enough to be studied and in this case specific SiC-coated carbon electrodes were used. The key point of these studies was to find a method to separate the matrix compounds from the fissile material in order to link the head to the core of the process (electrochemical separation or liquid–liquid reductive extraction in the case of a pyrochemical reprocessing).

Promising Pyrochemical Actinide/Lanthanide Separation Processes Using Aluminum

Thermodynamic calculations have shown that aluminum is the most promising metallic solvent or support for the separation of actinides (An) from lanthanides (Ln). In molten fluoride salt, the technique of reductive extraction is under development in which the separation is based on different distributions of An and Ln between the salt and metallic Al phases. In this process molten aluminum alloy acts as both a reductant and a solvent into which the actinides are selectively extracted. It was demonstrated that a one-stage reductive extraction process, using a concentrated solution, allows a recovery of more than 99.3% of Pu and Am. In addition excellent separation factors between Pu and Ln well above 103 were obtained. In molten chloride media similar separations are developed by constant current electrorefining between a metallic alloy fuel (U60Pu20-Zr10Am2Nd3.5Y0.5Ce0.5Gd0.5) and an Al solid cathode. In a series of demonstration experiments, almost 25 g of metallic fuel was reprocessed and actinides collected as An-Al alloys on the cathode. Analysis of the An-Al deposits confirmed that an excellent An/Ln separation (An/Ln mass ratio = 2400) had been obtained. These results show that Al is a very promising material to be used in pyrochemical reprocessing of actinides.

Behaviour of Rare Earth Elements in Molten Salts in Relation to Pyrochemical Reprocessing of Spent Nuclear Fuels

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CEMOS (Catalyst Enhanced Molten Salt Oxidation) for Complete and Continuous Pyrochemical Reprocessing of Spent Nuclear Fuel: An Overview of a Viable New Technology for Next Generation Nuclear Reactors

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Development of a Phosphate Ceramic as a Host for Halide-contaminated Plutonium Pyrochemical Reprocessing Wastes

AbstractThe presence of halide anions in four types of wastes arising from the pyrochemical reprocessing of plutonium required an immobilization process to be developed in which not only the actinide cations but also the halide anions were immobilized in a durable, leach resistant form. AWE has developed such a process using Ca3(PO4)2 as the host material. Successful trials of the process using actinide-doped Type I waste (essentially a chloride-based waste) were carried out at PNNL where the immobilization of the waste in a form resistant to aqueous leaching was confirmed. Normalized mass losses determined using a modified MCC-1 test at 40°C/28 days were 12 x 10-6 g.m-2 and 2.7 x 10-3 g.m-2 for Pu and Cl, respectively. Accelerated radiation-induced damage effects are being determined with specimens containing 238Pu. No changes in the crystalline lattice have been detected with XRD after the 239Pu equivalent of 400 years ageing. Confirmation of the process for Type II waste (an oxyhydroxide-based waste) is currently underway at PNNL.Differences in the ionic state of plutonium in the four types of waste have required different surrogates to be used. Samarium chloride was used successfully as a surrogate for both Pu(III) and Am(III) chlorides. Early investigations into the use of HfO2 as the surrogate for Pu(IV) oxide in Type II waste showed some apparent differences in the phase assemblages of the surrogate and actinide-based products. However XRD examination of the products at higher resolution has demonstrated there is no significant difference and that for this work HfO2 is a suitable surrogate for PUO2.

Immobilization of Simulant ILW Actinide Wastes Containing Halides: Effect of Process Parameters on Wasteform Properties

ABSTRACTThe applicability of a process, originally developed to immobilize a chloride-rich intermediate level waste (ILW) arising from the pyrochemical reprocessing of plutonium, to alternative ILW waste-streams containing a large variety of cations, present as oxides together with KCl and CaF2, has been investigated. Non-radioactive studies, using simulated wastes containing hafnium and samarium as surrogates for the actinides, have been performed to establish the optimum processing parameters for the manufacture of wasteforms suitable for long-term storage/disposal. It has been demonstrated that the wastes can be immobilized by calcining with either Ca3(PO4)2 or CaHPO4 as the host. In excess of 97.5 % of the chloride present in calcined wastes, containing up to 3.2 wt% halides as a mixture of F- and Cl-, is immobilized when using this process. Whereas in the original calcined waste, the Cl- (about 12 wt%) was immobilized as chlorapatite, Ca5(PO4)3Cl, and spodiosite, Ca2(PO4)Cl, in these current wastes fluorapatite is the predominant halide containing phase. Conversion of the immobilized waste powder into a monolithic wasteform was achieved by sintering the powder with the addition of a sodium aluminium phosphate based glass. Both the calcined powder and the sintered monolithic wasteforms showed excellent durability in aqueous solution.

Potential of thorium molten salt reactorsdetailed calculations and concept evolution with a view to large scale energy production

We discuss here the concept of Thorium Molten Salt Reactor dedicated to future nuclear energy production. The fuel of such reactors being liquid, it can be easily reprocessed to overcome neutronic limits. In the late sixties, the MSBR project showed that breeding is possible with thorium in a thermal spectrum, provided that an efficient pyrochemical reprocessing is added. With tools developed around the Monte Carlo MCNP code, we first re-evaluate the performance of a MSBR-like reference system with 232Th/ 233U fuel. We find an important reduction of inventories and induced radiotoxicities at equilibrium compared to other fuel cycles, with a doubling time of about thirty years. We then study how to start this interesting reference system with theplutonium from PWR spent fuel. Such a transition appears slow and difficult, since it is very sensitive to the fissile quality of the plutonium used. Deployment scenarios of 232Th/ 233U MSBR-like systems from the existing French PWRs demonstrate the advantage of an upstream 233U production in other reactors, allowing a direct start of the MSBR-like systems with 233U. This finally leads us to explore alternatives to some MSBR features, for energy production with 232Th/ 233U fuel from the start. We thus test different options, especially in terms of core neutronics optimization and reprocessing unit adaptation.

The immobilization of chloride-containing actinide waste in a calcium phosphate ceramic host: ageing studies

AbstractImmobilization of radioactive wastes arising from the pyrochemical reprocessing of Pu metal with calcium phosphate as the host material has been investigated. Initial studies in which non-active elements are substituted for the actinide elements present in the wastes demonstrated that effective immobilization of both the cations and chloride ions was achieved within the chlorapatite Ca5(PO4)3Cl, and spodiosite, Ca2(PO4)Cl, mineral phases produced during the solid state reaction between host and waste. Follow-up studies with radioactive simulated waste composed of CaCl2, PuCl3 and AmCl3have confirmed the viability of the process and the durability of the product in aqueous solution. More recently a study of the long-term radiological stability and chemical durability of the ceramic product has commenced in which accelerated ageing of the ceramic phases is being achieved through the replacement of the 239Pu (half life 24,360y) that is contained in the actual waste by the short lived isotope 238Pu (half life 86.4y). The preparation of non-active and active samples and the crystal phases formed in each system as determined with XRD are herein reported. Data are presented which show that there is no detectable change in the XRD pattern of the samples containing 238Pu after 532 days of ageing, corresponding to around 400 years of real time ageing employing 239Pu. The detailed effects of ageing on the crystal structure and on the chemical durability of samples in aqueous solution are reported.

Behavior of molybdenum in pyrochemical reprocessing: A spectroscopic study of the chlorination of molybdenum and its oxides in chloride melts

The high temperature reactions of molybdenum and its oxides with chlorine and hydrogen chloride in molten alkali metal chlorides were investigated between 400 and 700 °C. The melts studied were LiCl–KCl, NaCl–CsCl and NaCl–KCl and the reactions were followed by in situ electronic absorption spectroscopy measurements. In these melts Mo reacts with Cl 2 and initially produces MoCl 6 2− and then a mixture of Mo(III) and Mo(V) chlorocomplexes, the final proportion depending on the reaction conditions. The Mo(V) content can be removed as MoCl 5 from the melt under vacuum or be reduced to Mo(III) by Mo metal. The reaction of Mo when HCl gas is bubbled into alkali chloride melts yields only MoCl 6 3−. MoO 2 reacts in these melts with chlorine to form soluble MoOCl 5 2− and volatile MoO 2Cl 2. MoO 3 is soluble in chloride melts and then decomposes into the oxychloride MoO 2Cl 2, which sublimes or can be sparged from the melt, and molybdate. Pyrochemical reprocessing can thus be employed for molybdenum since, after various intermediates, the end-products are chloride melts containing chloro and oxychloro anions of molybdenum plus molybdate, and volatile chlorides and oxychlorides that can be readily separated off. The reactions were fastest in the NaCl–KCl melt. The X-ray diffraction pattern of MoO 2Cl 2 is reported for the first time.

Treatment of molten salt wastes by phosphate precipitation: removal of fission product elements after pyrochemical reprocessing of spent nuclear fuels in chloride melts

The removal of fission product elements from molten salt wastes arising from pyrochemical reprocessing of spent nuclear fuels has been investigated. The experiments were conducted in LiCl–KCl eutectic at 550 °C and NaCl–KCl equimolar mixture at 750 °C. The behavior of the following individual elements was investigated: Cs, Mg, Sr, Ba, lanthanides (La to Dy), Zr, Cr, Mo, Mn, Re (to simulate Tc), Fe, Ru, Ni, Cd, Bi and Te. Lithium and sodium phosphates were used as precipitants. The efficiency of the process and the composition of the solid phases formed depend on the melt composition. The distribution coefficients of these elements between chloride melts and precipitates were determined. Some volatile chlorides were produced and rhenium metal was formed by disproportionation. Lithium-free melts favor formation of double phosphates. Some experiments in melts containing several added fission product elements were also conducted to study possible co-precipitation reactions. Rare earth elements and zirconium can be removed from both the systems studied, but alkaline earth metal fission product elements (Sr and Ba) form precipitates only in NaCl–KCl based melts. Essentially the reverse behavior was found with magnesium. Some metals form oxide rather than phosphate precipitates and the behavior of certain elements is solvent dependent. Caesium cannot be removed completely from chloride melts by a phosphate precipitation technique.

Formation of lanthanide phosphates in molten salts and evaluation for nuclear waste treatment

The formation of phosphates of the lighter lanthanides (Ln = La, Ce, Pr, Nd, Sm, Eu, Gd, Tb and Dy) by reaction with alkali metal phosphates has been studied in a LiCl–KCl eutectic melt between 450 and 650 °C and in a NaCl–KCl equimolar melt at 750 °C and under an inert atmosphere. Alkali metal ortho-, meta- and pyrophosphate precursors have been employed. The reaction results in the formation of single lanthanide or double alkali metal–lanthanide orthophosphates, the former favoured in LiCl–KCl melts and the latter in NaCl–KCl melts. The mean crystallite size of precipitated phosphates was evaluated from X-ray powder diffraction patterns, and found to be within 300–450 A. Increasing the melt temperature results in increasing the crystallite size of phosphate phases. This technique offers an attractive means of removing lanthanide fission products from chloride melts arising from pyrochemical reprocessing of spent nuclear fuels. LiCl–KCl based melts are recommended because the bulk of phosphate waste precipitated is smaller since normal rather than double phosphates are formed therein.

Investigation on the electric properties in molten quaternary systems by MD simulation

For the pyrochemical reprocessing of spent metallic fuels in molten salt, it is of importance to estimate the enrichment degree of Cs. The molecular dynamics simulation has been carried out on molten quaternary systems (Li, Na, K, Cs)Cl at 625K and (Li, Na, K, Cs)F at 727K for the various compositions in order to investigate the electric properties, i.e., the electric conductivity, self-diffusion coefficient, self-exchange velocity and the relative differences in the internal cation mobilities of Cs in molten LiCl-NaCl-KCl eutectic mixtures and in the FLINAK melts. These results allow us to conclude that the electric conductivities, self-diffusion coefficients and self-exchange velocities of Li+, Na+, K+ and Cs+ with reference to Cl− and F− have almost similar tendencies for each composition. We found it possible to enrich at up to χCs = 0.38 in molten LiCl-NaCl-KCl eutectic as well as LiCl-KCl system and up to χCs = 0.42 in the FLINAK melts as well as in molten FLINA system. In addition, the sequence of the simulated electric conductivity in molten quaternary alkali chloride and fluoride systems was in a fair agreement with that of the current simulated self-exchange velocities and self-diffusion coefficients.

Investigation of the electronic and dynamic properties of sodium and caesium exchanged zeolite by MO and MD simulations

For the pyrochemical reprocessing of spent metallic fuels in molten salt baths it is of importance to investigate the electronic and dynamic properties of the negative elements like Cs in aluminosilicates framework. The molecular orbital simulation has been performed on three types of clusters and 4A-zeolite frameworks with exchangeable alkali-ions containing as significant fission products in order to estimate the geometry optimization, the vibrational frequency factors and the electric densities, etc. These quantum chemical results enable us to conclude that the most stable structure is consistent with the X-ray results. Moreover, the obtained infrared spectrum was reproduced by the experimental results. Furthermore, the molecular dynamics simulation for Na-A and Cs-A zeolites has been carried out at 673 K in order to investigate the dynamics of Na+ and Cs+ions in Na-A and Cs-A zeolite frameworks. These results revealed that Na I ion in β-cage was more stable than the other Na ions in Na-A zeolite and Cs I ion in α-cage was maintained stability in Cs-A zeolite in consideration of the self-diffusion coefficients.

Dissolution of uranium nitrides in LiCl-KCl eutectic melt

A combination of the nitride fuel and pyrochemical reprocessing has been chosen as a candidate for the fuel cycle of the minor actinides (MA) burner systems. One important issue is the feasibility of the recycled use of 15N-enriched nitrogen in the nitride pyrochemical process. Dissolution of spent nitride fuel with an oxidation agent has been considered to be a promising method for recovering 15N-enriched nitrogen. The behavior of the dissolution of uranium nitrides (UN, U2N3) and UNCI with CdCl2 in LiCl-KCl eutectic melt is investigated to show feasibility of the method for uranium containing nitride fuel. Most of nitrogen is recovered as N2 gas resulting from the reaction of uranium nitrides and CdCl2 above 550 °C, which is higher than that reported for lanthanide nitrides, surrogates of transplutonium nitrides, with similar conditions. Formation of intermediate compounds is observed with the temperature below 500°C. ZrN, a candidate of dilution material for MA burner fuel, is hard to be dissolved with CdCl2 in LiCl-KCl eutectic salt even at high temperature such as 800 °C.

A Spectroscopic Study of Uranium Species Formed in Chloride Melts

The chlorination of uranium metal or uranium oxides in chloride melts offers an acceptable process for the head-end of pyrochemical reprocessing of spent nuclear fuels. The reactions of uranium metal and ceramic uranium dioxide with chlorine and with hydrogen chloride were studied in the alkali metal chloride melts, NaCl-KCl at 973 K, NaCl-CsCl between 873 and 923 K, and LiCl-KCl at 873 K. The uranium species formed therein were characterised from their electronic absorption spectra measured in situ. The kinetic parameters of the reactions depend on melt composition, temperature and chlorinating agent used. The reaction of uranium dioxide with oxygen in the presence of alkali metal chlorides results in the formation of alkali metal uranates. A spectroscopic study, between 723 and 973 K, on their formation and their solutions was undertaken in LiCl, LiCl-KCl eutectic and NaCl-CsCl eutectic melts. The dissolution of uranium dioxide in LiCl-KCl eutectic at 923 K containing added aluminium trichloride in the presence of oxygen has also been investigated. In this case, the reaction leads to the formation of uranyl chloride species.

Investigation on the electrical properties of molten quaternary systems (Li, Na, K, Cs)Cl and (Li, Na, K, Cs)F by MD simulation

For the pyrochemical reprocessing of spent metallic fuels in molten salts, it is of importance to estimate the degree of enrichment of Cs. Molecular dynamics simulation has been carried out on molten quaternary systems (Li, Na, K, Cs)Cl at 625 K and (Li, Na, K, Cs)F at 727 K for the various compositions in order to investigate the electrical properties, i.e. the electrical conductivity, the self-diffusion coefficient, the self-exchange velocity and the relative differences in the internal cation mobilities of Cs in molten LiCl–NaCl–KCl eutectic mixtures and in the FLINAK melts. These results allow us to conclude that the electrical conductivities, the self-diffusion coefficients and the self-exchange velocities of Li +, Na +, K + and Cs + with reference to Cl − and F − show almost the same tendencies for each composition. We demonstrated that it was possible to enrich Cs up to x Cs=0.38 in molten LiCl–NaCl–KCl eutectic as well as in the LiCl–KCl system and up to x Cs=0.42 in the FLINAK melts as well as in the molten FLINA system. In addition, the sequence of the simulated electrical conductivity in molten quaternary alkali chloride and fluoride systems was in fair agreement with that of the current simulated self-exchange velocities and self-diffusion coefficients.