Pyrometallurgical Reprocessing Research Articles

Precious Metal Distributions Between Copper Matte and Slag at High P_{{{text{SO}}_{ 2} }} in WEEE Reprocessing

The distributions of precious metals (gold, silver, platinum, and palladium) between copper matte and silica-saturated FeOx-SiO2/FeOx-SiO2-Al2O3/FeOx-SiO2-Al2O3-CaO slags were investigated at 1300 °C and P_{{{text{SO}}_{ 2} }} = 0.5 atm. The experiments were carried out in silica crucibles under flowing CO-CO2-SO2-Ar gas atmosphere. The concentrations of precious metals in matte and slag were analyzed by Electron Probe X-ray Microanalysis and Laser Ablation-High-Resolution Inductively Coupled Plasma-Mass Spectrometry, respectively. The precious metal concentrations in matte and slag, as well as the distribution coefficients of precious metals between matte and slag, were displayed as a function of matte grade. The present results obtained at P_{{{text{SO}}_{ 2} }} of 0.5 atm were compared with previous results at P_{{{text{SO}}_{ 2} }} of 0.1 atm for revealing the effects of P_{{{text{SO}}_{ 2} }} and selected slag modifiers (CaO and Al2O3) on precious metal distributions at copper matte smelting conditions. The present results also contribute experimental thermodynamic data of precious metal distributions in pyrometallurgical reprocessing of electronic waste via copper smelting processes.

Removal of rare-earth fission products from molten chloride salt used in pyroprocessing by precipitation for consolidation into glass-bonded sodalite waste form

As an alternative method to absorption into zeolite-A for removing rare-earth (RE) fission products (FPs) accumulated in spent molten chloride salt generated by the pyrometallurgical reprocessing of spent nuclear fuel, the precipitation of the RE FPs followed by consolidation into a glass-bonded sodalite waste form has been proposed to reduce the waste volume. Chemical forms of the precipitates generated by the addition of Li2O or Li2CO3 in LiCl-KCl melts containing chlorides of REs were identified. Then, the salt adhering to the Gd2O3 precipitate, which was generated by the addition of Li2O into LiCl-KCl molten salt containing GdCl3, was removed by vacuum distillation. The distillation product or its simulant (Gd2O3) was mixed with zeolite-A and fresh zeolite, and then consolidated into glass-bonded sodalite under uniaxial pressing at the maximum temperature and pressure of 915 °C and 1.3 × 105 Pa, respectively. Examinations of the consolidated products showed that a dense product was fabricated and that gadolinium existed as NaGd9Si6O26, as expected from the literature, or Gd2O3 in the products.

Effect of reduced dehalogenation on the performance of Y zeolite-based sintered waste forms

The properties of a dehalogenated salt waste form (DSWF) were investigated for the immobilization of electrorefiner salt waste from the pyrometallurgical reprocessing of used nuclear fuel. Dehalogenation of salt through reaction with hydrogen-Y zeolite provides a viable means of decreasing solid waste volume prior to immobilization. The presence of residual salt (from insufficient dehalogenation reaction time or inefficient operation) on waste form properties is also evaluated via hot stage X-ray diffraction and leach testing. Residual salt appears to volatilize during sintering, producing a secondary waste stream. The kinetic dissolution rates of DSWFs show evidence of congruent dissolution and exhibit lower dissolution rates compared to the baseline GBS advanced ceramic waste form. Additional characterization is required to determine behavior due to saturation effects.

Synthesis and characterization of sintered H–Y zeolite-derived waste forms for dehalogenated electrorefiner salt

A novel nuclear waste form was synthesized for the disposal of electrorefiner salt from pyrometallurgical reprocessing of used nuclear fuel. Its synthesis constitutes two steps: waste salt dehalogenation via ion exchange with H–Y zeolite, followed by thermal treatment for phase transition and consolidation. Salt cation-loaded zeolite exhibiting greater than 90% dehalogenation was sintered for 12 h at 925 °C and achieved 3 vol% open porosity (determined via Archimedes method). Microstructural imaging and structural characterization revealed a multiphase assemblage of alkali aluminosilicate and mixed oxide phases. Chemical durability experiments were performed using modified ASTM C1308 tests in demineralized water and different silica solutions. Si-containing species appear to exhibit congruent dissolution, while the release behaviors of alkali elements appear to include diffusion-limited contributions. The 4-day cumulative releases for the dehalogenated salt waste form (DSWF) are less than 33% of those measured for the baseline glass-bonded sodalite advanced ceramic waste form (GBS ACWF). Furthermore, the dehalogenation and consolidation processing steps provide a 35% decrease in required storage volume for a given waste salt mass relative to the GBS ACWF.

Postirradiation examination on metallic fuel in the AFC-2 irradiation test series

Irradiation experiments on metallic fuel alloys containing rare earth additions to simulate potential carry-over from pyro-metallurgical reprocessing and minor actinide additions were performed in the Idaho National Laboratory Advanced Test Reactor. These tests were designated AFC-2A and AFC-2B. Additionally irradiation experiments were also performed on metallic fuel alloys containing minor actinide additions that were fabricated by different techniques. These tests were designated AFC-2E. These pins were removed from the reactor after achieving burnups between 6 and 11% fission per initial heave metal atom. Examination of these pins revealed that they were unintentionally operated beyond their desired power and temperature conditions. Highlights from the postirradiation examination of these experiments is discussed. A variety of issue contributed to the poor performance of these irradiation tests including capsule design, reactor power variations and fabrication issues. While all the issues contributed, the major issue was determined to be unanticipated variations in reactor power. Through comparisons between these experiments and historical off-normal experiments, it was possible to conclude bound the operating temperatures of these experiments. There likely was some in-pile melting in the rare earth addition experiments (AFC-2A, AFC-2B), but it is not likely that there was in-pile melting in the fabrication technique experiments (AFC-2E).

Parameter Surveys on Glass-Bonded Sodalite Synthesis Conditions from Spent Salt Generated in Pyroprocess

A parameter-based survey of the synthesis conditions by a so-called pressureless consolidation method to fabricate glass-bonded sodalite waste form for stabilizing fission products generated in pyrometallurgical reprocessing of spent metal fuel was performed. The maximum temperature, the heating duration at the maximum temperature, the glass fraction in the initial material, and the weight load used for pressing the material were chosen as the variable parameters. Accordingly, modified conditions to reduce the maximum temperature and increase the weight load were selected for reducing the volatilized-salt ratio during the heating and the free-salt ratio in the product. By fabricating a simulated waste under the modified conditions, the effect of changing the conditions was confirmed. Leaching tests in pure water using the consolidated products fabricated under both reference and modified conditions showed that the stability of the products was not significantly deteriorated by modifying the heating conditions.

Direct separation of uranium from lanthanides (La, Nd, Ce, Sm) in oxide mixture in LiCl-KCl eutectic melt

This paper reports the direct separation of U from lanthanides (Ln) in oxide mixtures in LiCl-KCl eutectic, a method that could simplify the pyrometallurgical reprocessing of UO2-based spent fuels. In the melt, the mixture of UO2 and lanthanide oxides (Ln2O3/LnO2) was chloridized by AlCl3, and then U was selectively recovered by forming U-Al alloys through the redox reactions of U(IV)/U(III) and U(III)/AlxU. The results showed that U could be efficiently separated from lanthanides with high separation factors (SFU/La = 1.1 × 104, SFU/Ce = 1.6 × 105, and SFU/Sm = 2.5 × 104). X-ray diffraction analysis indicated the predominant separation products to be Al3U and Al4U, whereas the real extraction rate of U was probably affected by trace amounts of O2− in the melt. For the first time, single-cell operation to separate U from Ln in mixed oxides is shown to be feasible by using AlCl3 as chlorination agent followed by electrodeposition.

Molten Salt Assisted Electrochemical Separation of Spent Fuel Surrogates by Partial Direct Reduction and Selective Anodic Dissolution

Pyrometallurgical reprocessing of spent nuclear fuels has experienced growing interest in recent years as it presents a desirable and proliferation resistant separation process capable of significantly reducing volumes of high level waste (HLW). Additionally, the use of molten salts in pyro-reprocessing is seen as forward compatible with the new GEN IV fast neutron nuclear reactors. In this work, the ability to selectively partition spent fuel surrogate species in solid phase by molten salt electrochemistry was assessed in a proof of concept experiment by the use of the 2NiO-CeO2 binary. In a three electrode cell using potentiostatic electrolysis it was found that close to 80% of the Ni presence in the surface of the precursor was removed following a partial direct reduction of NiO by the FFC Cambridge Process, followed by anodic dissolution to remove the metallic Ni phase. The mixture of more analogous electrochemical surrogates ZrO2 and CeO2 to that of UO2 and PuO2 were then assessed to further confirm the application of this novel recycling technology.

Development of Treatment Process for Anode Residue from Molten Salt Electrorefining of Spent Metallic Fast Reactor Fuel

A new treatment process was proposed for the anode residue from a molten salt electrorefining step in the pyrometallurgical reprocessing of spent metallic fast reactor fuel. This treatment process consists of two steps: (a) oxidation of the remaining actinides in the anode residue by the addition of CdCl2 and (b) removal of the accompanying chloride by high-temperature distillation. The oxidation of the remaining uranium by CdCl2 was studied using anode residue from previous electrorefining experiments using U-Zr alloys. The reaction between uranium and CdCl2 was completed in [approximately]2 days with a satisfactory chlorine balance among the species in the molten chlorides solvent. A high uranium oxidation rate was attained by appropriately controlling the rate of CdCl2 addition. The high-temperature distillation tests were carried out at 1473 K with pressure of [approximately]300 Pa to remove the solvent accompanying the anode residue. The chloride content in the anode residue was lowered to 1% to 2.5% by the distillation operation. Although the anode residue was heated to 1673 to 1773 K at a pressure of [approximately]50 kPa after the distillation, it was not melted completely. The remaining ratio of uranium after the electrorefining and the above treatment process was evaluated to be 0.04% to 0.20%. Material flow calculations were performed for a pyrometallurgical reprocessing facility equipped with the anode residue treatment process. It showed that (a) the chlorine and uranium supply/demand balance is maintained unless the remaining ratio of uranium after electrorefining exceeds a certain value and (b) the addition of the anode residue treatment process does not have an adverse effect on either the performance of the overall process or the facility design.

Development of Transport Technologies for High-temperature Fluid in Pyrometallurgical Reprocessing

The development of the engineering technology necessary for pyrometallurgical reprocessing is a key issue for industrial realization. The development of high-temperature transport technologies for molten salt and liquid cadmium is crucial for pyrometallurgical processing; however, there have been very few transport studies on high- temperature fluids. In CRIEPI, the transport technologies of molten salt and liquid cadmium have been developed for several types of transports methods. Based on the results of transport tests, engineering scale process equipments connected each other with the transport technology have been developed, and installed in the Ar atmosphere glove box for integrated tests.

Separation of SmCl3 from SmCl3-DyCl3 system by Electrolysis in KCl-LiCl-MgCl2 Molten Salts

Pyrometallurgical reprocessing of spent nuclear fuel was now considered as one of the most promising options for an innovative nuclear fuel cycle. In this paper, the electrochemical behavior of Sm(III) and Dy(III) was studied by cyclic voltammetry in KCl-LiCl molten salts. Addition of magnesium chloride resulted in positive deposition potential of Dy(III), lower electrolysis temperature and higher recovery ratio of rare earth. According to different deposition potentials of Sm(III) and Dy(III), electrolysis in the molten KCl-LiCl-MgCl2 system constituted the main step in this reprocess, where SmCl3 was separated from the SmCl3-DyCl3 system at -2.20 V (vs. Ag/AgCl reference electrode) and the DyCl3 were formed the Dy-Mg alloys. It was proved that the main structures of the obtained alloys were Dy and DyMg3 without Sm by the analysis of XRD of alloys.

Electrochemical Measurement of Diffusion Coefficient of Actinides and Rare Earths in Liquid Cd

Liquid Cd is used as a reaction medium in the pyrometallurgical reprocessing for spent nuclear fuels. The diffusion coefficient of a series of key elements, such as actinides and rare earths, in liquid Cd is one of the important physical properties for a kinetic analysis of the pyrometallurgical reprocessing. Recently, the authors applied an electrochemical method to the measurement of a diffusion coefficient in liquid metal. This presentation summarizes the diffusion coefficients of actinides (U and Pu) and rare earths (La, Pr, Nd, Gd, Y and Sc) in liquid Cd measured by the electrochemical method.

Studies on Novel Matrices for High Level Waste from Fast Reactor Fuel Reprocessing

Abstract High level waste (HLW) originating from aqueous reprocessing of fast reactor fuels consists mainly of fission products, minor actinides and noble metals which are higher in concentration compared to that from thermal reactors. Borosilicate glass may not be an efficient matrix to fix this waste. Similarly, chloride HLW arising from the pyro-metallurgical reprocessing has poor solubility in borosilicate glass. The present study explores various alternate systems consisting of a glass (iron phosphate glass, IPG), ceramic (monazite) and glass -ceramic (apatite) for the immobilization of such HLW. Studies related to synthesis, thermal stability and waste loading, point to the potential of the three matrices to emerge as viable waste forms for different types of HLW

金属電解法乾式再処理プロセス実用化研究開発 (12) 液体カドミウム中のPuの拡散係数の電気化学的測定

金属電解法乾式再処理プロセス実用化研究開発 (12) 液体カドミウム中のPuの拡散係数の電気化学的測定

Countercurrent Extraction Test with Continuous Flow of Molten LiCl-KCl Salt and Liquid Cd for Pyro-Reprocessing of Metal FBR Fuel

The three-stage countercurrent extraction test with molten LiCl-KCl and liquid Cd was carried out for the development of spent salt treatment process in pyrometallurgical reprocessing of metal FBR fuel. Both salt and Cd solvent were supplied successively from each feed tank into the extractor at a constant flow rate and recovered separately into individual recovery tanks. Uranium, transuranic elements, and rareearth elements were substituted by Ce, Gd, and Y, respectively. More than 98% of Ce and 93% of Gd were recovered, and 27% of Y was corecovered in the Cd solvent. The separation factors of Gd and Y against Ce were evaluated to be 0.28–0.29 and 0.0057–0.0073, respectively. The former is almost equal to and the latter is better than that in the equilibrium. As the connecting part between each stage had a simple structure, a small amount of Cd in each stage mixed with that in the neighboring stage, which made the separation efficiency lower than expected. A calculation study showed the relationship between the Cd mixing rate and recovery yield or separation factor, and it showed that the recovery yield and separation efficiency were expected to be much improved by the modification of the connecting part to prevent Cd from mixing.

Development of High-Temperature Transport Technologies of Molten Salt Slurry in Pyrometallurgical Reprocessing

Development of High-Temperature Transport Technologies of Molten Salt Slurry in Pyrometallurgical Reprocessing

Development of an Innovative Electrorefiner for High Uranium Recovery Rate from Metal Fast Reactor Fuels

To increase the uranium recovery rate of molten salt electrorefining step in pyrometallurgical reprocessing of metallic fast reactor fuels, tests were carried out using electrorefiners equipped with mechanisms for scraping cathode deposits. After the modifications in the design of the anode basket and scraper mechanism, no stalling of the anode and scraper rotation due to interference by cathode deposits occurred. Under the condition that codissolution of zirconium and uranium was allowed in order to obtain maximum throughput, a current of 400–450A was maintained until 82% of the initially loaded uranium was recovered. The uranium recovery rate for the same duration reached 789 g U/h (32.9 g U/h_L per electrode volume). On the assumption that an electrorefiner operates for 20 h/d and 200 d/y in an actual pyrometallugical reprocessing facility, this result corresponds to a uranium recovery rate of 3.16 t U/y using only one electrode assembly of about 30 cm diameter, which should be a sufficiently high performance for practical use. From these results, the engineering feasibility of uranium recovery using an electrorefiner with cathode deposit scraper mechanism has been demonstrated.

Development of High-Temperature Transport Technologies for Liquid Cadmium in Pyrometallurgical Reprocessing

Pyrometallurgical reprocessing is one of the most promising technologies for the advanced fuel cycle with favorable economic potential and intrinsic proliferation-resistance. The feasibility of pyrometallurgical reprocessing has been studied through many laboratory-scale experiments. Hence the development of the engineering technology necessary for pyrometallurgical reprocessing is a key issue for its industrialization. The development of high-temperature transport technologies for molten salt and liquid cadmium is crucial for pyrometallurgical processing; however, there have been a few transport studies on high-temperature fluids. In this study, a metal transport test rig was installed in an argon glove box with the aim of developing technologies for transporting liquid cadmium at approximately 773 K. The transport of liquid Cd using gravity was controlled by adjusting the valve. The liquid Cd was transported by a suction pump against a 0.93 m head and the transport amount of Cd was well controlled with the Cd amount and the position of the suction tube. The transportation of liquid cadmium at approximately 700 K could be controlled at a rate of 0.5–2.5 dm3/min against a 1.6 m head using a centrifugal pump.

Development of High Temperature Molten Salt Transport Technology for Pyrometallurgical Reprocessing

Pyrometallurgical reprocessing technology is currently being focused in many countries for closing actinide fuel cycle because of its favorable economic potential and an intrinsic proliferation-resistant feature due to the inherent difficulty of extracting weapons-usable plutonium. The feasibility of pyrometallurgical reprocessing has been demonstrated through many laboratory scale experiments. Hence the development of the engineering technology necessary for pyrometallurgical reprocessing is a key issue for industrial realization. The development of high-temperature transport technologies for molten salt and liquid cadmium is crucial for pyrometallurgical processing; however, there have been very few transport studies on high-temperature fluids. In this study, a salt transport test rig was installed in an argon glove box with the aim of developing technologies for transporting molten salt at approximately 773 K. The gravitation transport of the molten salt at approximately 773 K could be well controlled at a velocity from 0.1 to 1.2 m/s by adjusting the valve. Consequently, the flow in the molten salt can be controlled from laminar flow to turbulent flow. It was demonstrated that; using a centrifugal pump, molten salt at approximately 773 K could be transported at a controlled rate from 2.5 to 8 dm3/min against a 1 m head.

Single-Stage Extraction Test with Continuous Flow of Molten LiCl-KCl Salt and Liquid Cd for Pyro-Reprocessing of Metal FBR Fuel

A single-stage extraction test with the continuous flow of both molten salt and liquid Cd was carried out as a preliminary step of the development of a multistage countercurrent extraction process for the pyrometallurgical reprocessing of metal FBR fuel. Ce, Gd, and Y were selected as the substitutes for U, other actinides, and rare-earth elements, respectively. More than 97% of Ce could be recovered in one of the experiments, in which 97% of Gd and 63% of Y were also recovered. This meant that high recovery yield could be obtained in the system, but mutual separation was not sufficient when the high recovery yield was required in a single-stage extraction. In another experiment in which the recovery yields of Ce, Gd, and Y were about 87, 70, and 8%, respectively, the separation factors of Gd and Y against Ce were 0.27 and 0.0092, respectively, which were about the same as their equilibrium values. The result showed that higher separation efficiency was achieved under the condition of relatively lower recovery yield. The mass transfer coefficient of rare-earth elements in the salt was estimated to be 0.14 cm/s, which was much larger than that in the previous experiments without a baffle plate in the extractor.