Search

The management of radioactive carbon (C-14) from spent nuclear fuel (SNF) in a voloxidation process is vital to prevent radioactive contamination of the environment. Thus, a double alkali method was applied to absorb the gaseous phase of C-14 (CO2) and to immobilize radioactive carbon into a stable structure. Based on the two-film theory, mass transfer and enhancement factor were evaluated for CO2 absorption in NaOH solution with regards to the effects of operating conditions such as the solution concentration, CO2 partial pressure, and gas flow rate on the absorbing performance. The absorption tests were carried out targeting the successful capture of CO2 released from SNF with a high decontamination factors (DF) more than 103. Causticization with Ca(OH)2 leads to the immobilization of absorbed carbon into a scalenohedral calcite (CaCO3) crystal, and its stable and nonporous characteristics suggested that calcite is a suitable structure for preparing waste forms stored in a geological repository.

Management of radioactive 129 I is a challenging issue due to its significantly long half-life and negative health effects. In this study, silver iodide (AgI)-incorporated silver phosphate glasses were prepared to immobilize 129 I for a long time. A melt-quenching method was employed to synthesize the glasses at 450 °C to prevent the evolution of I during the heat treatment. Characterization of the AgI-Ag 2 O-P 2 O 5 specimens revealed that the addition of AgI was possible up to 60 mol% into the glassy matrix although there were some amounts of Ag 4 I(PO 4 ) and AgI clusters. This result was supported by the structural analysis which showed that the major iodine environment was composed of Ag and O regardless of the specimen compositions. The normalized leaching rates obtained by the 7-day PCT-A test showed that the chemical stability of iodine was satisfactory with the lowest leaching rate of 1.7 × 10 −4 g m −2 /d. Our findings in this study demonstrate the potential of silver phosphate glass for the immobilization of radioactive 129 I.

Al2O3-containing silver phosphate glasses were synthesized to investigate the feasibility of phosphate glasses for the immobilization of radioactive iodine (129I) present in spent nuclear fuel. Characterizations were performed by X-ray diffraction, Fourier transformed infrared spectroscopy, and scanning electron microscopy coupled with energy dispersive spectroscopy to examine structures, bonding properties, surface morphology, and elemental distribution of the synthesized glasses. The principal results showed that iodine became more strongly immobilized in the phosphate glasses with the addition of Al2O3, which was confirmed by the decrease of iodine leaching rates with approximately one order of magnitude. The present study would be helpful to decide whether Al2O3-containing silver phosphate glasses could be used as a candidate matrix to incorporate 129I.

In this study, the thermal detachment of rare earth (RE)-rich particles adhered to RE-poor particles separated by the high-temperature oxidation of (U,RE)3O8 powder was investigated by repeated oxidation after reduction at low temperature using air and ozone as oxidizing agents. The RE-rich particles detached from RE-poor particles were collected using a magnet of 0.8 T. The relative degree for the detachment and collection of RE-rich particles from RE-poor particles was evaluated by the observation of particle morphologies by SEM. As most of the RE-poor U3O8 particles during oxidation at low temperature were broken into plates, the detachment of RE-rich particles from RE-poor particles increased with increasing RE-rich particle sizes dependent on the oxidation temperature of (U,RE)3O8 powder. The detachment of the RE-rich particles increased with decreasing oxidation temperature from 500 to 360 °C under air. Under ozone, the (U,RE)O2 pellets, RE-rich and poor UO2 particles, and RE-rich U4O9 particles were not oxidized, while the RE-poor U3O8 particles were converted to the RE-poor UO3 particles with very low magnetic susceptibility. The collection of RE-rich particles increased with increasing magnetic susceptibility difference between the RE-rich and RE-poor particles.

The electrolytic reduction rate of porous UO2 pellets in a LiCl salt was investigated for various applied charges. The degree of reduction (α) value was evaluated from the ratios of cross-sectional areas of the reduced and oxide parts. An analysis of the experimental results revealed that the first-order reaction model is the best geometry function to describe the reduction reaction. An electrolytic reduction rate equation was proposed using the first-order model, although it was available in a limited region of (0≤α≤0.56). A power law based reaction rate equation was also suggested for the whole range of α, and the reaction time for a complete reduction, estimated using the power law equation, was confirmed through the experimental results. Changes in the Li-Li2O concentration around the reduced pellets for various applied charges were also measured, which increased up to 23 wt% with increasing α.

The pyrochemical process, which uses a dry method to recycle used nuclear fuel generates waste LiCl–KCl salt containing radioactive lanthanide elements. To reuse LiCl–KCl salt, the lanthanide elements are separated through a precipitation method promoted by oxygen sparging and the separated fission product of lanthanide oxide should be fabricated into durable wasteforms sustainable for several 1,000 years to store in a final geological repository. Herein, we report the fabrication of a borosilicate glass based wasteform with a glass matrix of SiO2–Al2O3–B2O3 having a high waste loading of 50 wt% lanthanide oxide. Th physical properties of four kinds of wasteforms having a different lanthanide oxide waste composition were evaluated. To investigate the long-term physical stability of each sample having 50 wt% lanthanide oxide waste loading, time–temperature–transformation (TTT) test was conducted at 500 and 700 °C for 60 and 180 h, and the physical properties were evaluated after each TTT test.

A study on an optimal method for the phosphorylation of rare earth chlorides in LiCl–KCl eutectic waste salt generated the pyrochemical process of spent nuclear fuel was performed. A reactor with a pitched four blade impeller was designed to create a homogeneous mixing zone in LiCl–KCl eutectic salt. A phosphorylation test of NdCl 3 in the salt was carried out by changing the operation conditions (operation temperature, stirring rate, agent injection amount). Based on the results of the test, a proper operation condition (450 °C, 300 rpm, 1 eq. of phosphorylation agent) for over a 0.99 conversion ratio of NdCl 3 to NdPO 4 was determined. Under this condition, multi-component rare earth (La, Ce, Pr, Nd, Sm, Eu, Gd, Y) chlorides were effectively converted into phosphate forms. It was confirmed that the existing regeneration process of LiCl–KCl eutectic waste salt can be greatly improved and simplified through these phosphorylation test results.

A minimization of waste salt is one of the most important issues for the optimization of pyroprocessing. The separation of fission products in waste salts and the reuse of purified waste salt are promising strategies for minimizing the waste salt amounts. The phosphate precipitation of lanthanide is currently being considered for eutectic (LiCl–KCl) waste salt purification. In this research, the effects of molten salt temperature (400–550 °C) and reaction time (max. 180 min) upon conversion into the phosphate of lanthanides was investigated using 1 and 3 kg of eutectic salt. The conversion efficiency of lanthanides to molten salt-insoluble precipitates and phosphates was increased with an increase in molten salt temperature and operating time until it attained a specific temperature and time. K3PO4 as a precipitant was more favorable than Li3PO4 in terms of reactivity. To obtain over a 99% overall conversion efficiency, about 30 min was required in the case of using K3PO4 at 450 °C, but about 120 min in t...

The separation characteristics of surrogate rare-earth fission products in a eutectic (LiCl–KCl) molten salt were investigated. This system is based on the eutectic salt used for the pyroprocessing treatment of used nuclear fuel (UNF). The investigation was performed using an integrated rare-earth separation apparatus comprising a precipitation reactor, a solid detachment device, and a layer separation device. To separate rare-earth fission products, a phosphate precipitation method using both Li3PO4 and K3PO4 as a precipitant was performed. The use of an equivalent phosphate precipitant composed of 0.408 molar ratio-K3PO4 and 0.592 molar ratio-Li3PO4 can preserve the original eutectic ratio, LiCl-0.592 molar ratio (or 45.2 wt%), as well as provide a high separation efficiency of over 99.5% under conditions of 550 °C and Ar sparging when using La, Nd, Ce, and Pr chlorides. The mixture of La, Nd, Ce, and Pr phosphate had a typical monoclinic (or monazite) structure, which has been proposed as a reliable host matrix for the permanent disposal of a high-level waste form. To maximize the reusability of purified eutectic waste salt after rare-earth separation, the successive rare-earth separation process, which uses both phosphate precipitation and an oxygen sparging method, were introduced and tested with eight rare-earth (Y, La, Ce, Pr, Nd, Sm, Eu and Gd) chlorides. In the successive rare-earth separation process, the phosphate reaction was terminated within 1 h at 550 °C, and a 4–8 h oxygen sparging time were required to obtain over a 99% separation efficiency at 700–750 °C. The mixture of rare-earth precipitates separated by the successive rare-earth separation process was found to be phosphate, oxychloride, and oxide. Through the successive rare-earth separation process, the eutectic ratio of purified salt maintained its original value, and impurity content including the residual precipitant of purified salt can be minimized.

The pyroprocessing which uses a dry method to recycle spent oxide fuel generates a waste LiCl salt containing radioactive elements. To reuse LiCl salt, the radioactive impurities has to be separated by the purification process such as layer-melt crystallization. To enhance impurity separation efficiency, it is important to understand the inclusion mechanism of impurities within the LiCl crystal. Herein, we report the inclusion properties of impurities in LiCl crystals. First of all, the substitution enthalpies of Cs + , Sr 2+ , and Ba 2+ impurities with 0–6 at% in LiCl crystal were evaluated via first-principles calculations. Also, the molten LiCl containing 1 mol of Cs + , Sr 2+ , and Ba 2+ impurities was crystallized through the experimental layer-melt crystallization method. These substitution enthalpy and experiment clarify that a high substitution enthalpy should result in the high separation efficiency for an impurity. Furthermore, we find that the electron density map gives a clue to the mechanism for inclusion of impurities into LiCl crystal.

Radioactive rare earth chlorides in waste LiCl–KCl molten salts have to be separated as a stable form to minimize waste volume and to achieve stable solidification. In this work, thermal behavior of rare earth chlorides (CeCl3, GdCl3, NdCl3, PrCl3) was investigated in an oxygen condition to recover rare earth oxides from a LiCl–KCl–RECl3 system. The rare earth chlorides in the LiCl–KCl molten salts were smoothly converted to an oxychloride form at a higher temperature than 650 °C, except for CeCl3. CeCl3 was totally converted to an oxide from at a higher temperature than 450 °C. The rare earth oxychlorides (GdOCl, NdOCl, PrOCl) were effectively converted to oxide forms at a higher temperature than 1100 °C. It was confirmed that rare earth oxides can be recovered from a LiCl–KCl–RECl3 system without impurity generation.

The effects of operating conditions such as crystal growing rate and initial impu rity concentration on separation(or concentration) of cesiu m and strontium fission products in the LiCl waste salt generated fro m an electro lytic reduction process of spent oxides on lab-scale layer melt crystallization process were analyzed. In a layer crystallizat ion process, impurities(CsCl and SrCl2) are concentrated in a s mall part of the LiCl salt due to the difference in the solubility in the melt phase and the solid l phase. The crystal growth rate strongly affects the crystal structure, therefore the separation efficiency, but the effect of the in itial Cs and Sr concentrations in LiCl mo lten salts was nearly negligible within the experimental range.

The Korea Atomic Energy Research Institute (KAERI) has been developing pyroprocessing technology for recycling useful resources from spent fuel since 1997. The process includes pretreatment, electroreduction, electrorefining, electrowinning, and a waste salt treatment system. This paper briefly addresses unit processes and related innovative technologies. As for the electroreduction step, a stainless steel mesh basket was applied for adaption of granules of uranium oxide. This basket was designed for ready handling and transfer of feed material. A graphite cathode was used for the continuous collection of uranium dendrite in the electrorefining system. This enhances the throughput of the electrorefiner. A particular mesh type stirrer was designed to inhibit uranium spill-over at the liquid Cd crucible. A residual actinide recovery system was also tested to recover TRU tracer. In order to reduce the waste volume, a crystallization method is employed for Cs and Sr removal. Experiments on the unit processes were tested successfully, and based on the results, engineering-scale equipment has been designed for the PRIDE (PyRoprocess Integrated inactive DEmonstration facility).

A lab-scale lanthanide precipitation apparatus, which has a 4 kg/batch size, was installed and tested. By using this lab-scale apparatus, cooxidative precipitation experiments of lanthanides were carried out. As lanthanides, 8 rare-earth elements (Y, La, Ce, Pr, Nd, Sm, Eu, and Gd) were used. By a reaction with oxygen, La, Pr, Nd, Sm, Eu, and Gd were converted to their oxychloride form (REOCl) and Ce, Pr, and Y to their oxide form (REO2, RE2O3). Since these lanthanide oxides or oxychlorides are nearly insoluble to eutectic salt, they were all precipitated by a free settling at the bottom, where about 7 h of precipitation time was required. It was found that under the conditions of 750°C salt temperature, a 12 h sparging time, and 5 L/min oxygen sparging rate, all the used rare-earth elements showed over 99% oxidation efficiency. However, in the case of 800°C temperature, they show over 99% conversion efficiencyonly after 6 h.

This paper is a study on a stable destruction method of radioactive waste ion exchange resins. According to the resin TGA results, its decomposition occurred through three stages. And a sufficient retention time of the resins and an effective retention capacity of SO2 gases and the doped metal compounds were required to destruct resins doped with radioactive metals stably. The resins doped with radioactive metal surrogates were effectively destructed in the lab-scale MCO system. CO and SO2 emissions were below 100 and 1 ppm, respectively. And the surrogates were collected more than 99.9% in the molten carbonate. Thus, the resins can be destructed stably in the MCO process.

Molten salt oxidation process is one of the most promising alternatives to incineration that can be used to effectively destroy the organic components of mixed and hazardous wastes. To detect the flow characteristics of the molten salt oxidation process (air-molten carbonate salt two-phase flow), differential pressure fluctuation signals from a molten salt oxidation process have been analyzed by adopting the stochastic methods. Effects of the input air flow rate (0.05-0.22 m/sec) and the molten salt temperature (870-970 o C) on the phase holdup and flow characteristics are studied. The gas holdup increases with an increasing molten salt temperature due to the decrease of the viscosity and surface tension of the molten carbonate salt. It is found that a stochastic analysis of the differential pressure signals enables us to obtain the flow characteristics in the molten salt oxidation process. The experimentally obtained gas hold- up data in the molten salt reactor were well described and characterized by means of the drift-flux model.

For the reuse of a waste salt from an electrorefining process of a spent oxide fuel, a separation of rare earth elements by an oxidative precipitation in a LiCl-KCl molten salt was tested without using precipitate agents. From the results obtained from the thermochemical calculations by HSC Chemistry software, the most stable rare earth compounds in the oxygen-used rare earth chlorides system were oxychlorides (EuOCl, NdOCl, PrOCl) and oxides (CeO 2 , PrO 2 ), which coincide well with results of the Gibbs free energy of the reaction. In this study, similar to the thermochemical results, regardless of the sparging time and molten salt temperature, oxychlorides and oxides were formed as a precipitant by a reaction with oxygen. The structure of the rare earth precipitates was divided into two shapes: small cubic (oxide) and large plate-like (tetragonal) structures. The conversion efficiencies of the rare earth elements to their molten salt-insoluble precipitates were increased with the sparging time and temperature, and Ce showed the best reactivity. In the conditions of 650 °C of the molten salt temperature and 420 min of the sparging time, the final conversion efficiencies were over 99.9% for all the investigated rare earth chlorides.

In this work, a new approach to remove fission products including decay heat elements was proposed. This study aims at providing a new way to minimize the amount of waste salt for a repository, while removing the high decay heat fission products [Cs, Sr, Ba, and Y including other rare earth (RE) elements] from the waste salts generated during a chloride pyroprocessing procedure. These elements were removed in consecutive order from the pyroprocessing units. First, Cs could be released in the form of an oxide gas during voloxidation of UO 2 and captured by a fly-ash filter. Then, Sr was recovered in the form of carbonate precipitates from the LiCl waste salt generated during the course of an electoreduction process, by using Li 2 CO 3 . Finally, RE elements plus yttrium in the spent LiCl-KCl waste salt generated during electrorefining were removed in the form of oxides (or oxychlorides) by using an oxygen sparging method. It was confirmed that the removal yields of each element were ∼90% for Cs at ∼1473 K, >99% for Sr at a molar ratio of [Li 2 CO 3 /SrCl 2 = 3], and >99% for the RE elements plus yttrium. Using these successes as a basis, a reference flow sheet for removing the high decay heat elements from pyroprocessing units is presented in this work. Also, a salt regeneration system to minimize the amount of waste salt is proposed in this study.

The carbonate reaction of some alkaline-earth chlorides was investigated by a carbonate agent injection method in LiCl-KCl eutectic salts containing both SrCl2 and BaCl2 and LiCl molten salts containing SrCl2. The effects of the injected molar ratio of a carbonate agent (Li2CO3 or K2CO3) and the temperature (450–750°C) on the conversion efficiencies of the strontium and barium chloride to their carbonates were determined. The forms of strontium and barium carbonate resulting from the carbonate reaction with carbonate agents were identified by XRD and SEM-EDS analyses. In these experiments, the carbonate agent injection method can carbonate strontium and barium chlorides effectively at over 99% under LiCl-KCl eutectic and LiCl molten salt conditions. For LiCl-KCl eutectic molten salts, carbonation efficiency was more favorable in the case of K2CO3 injection than in the case of Li2CO injection, where strontium and barium were carbonated in the form of Ba0.5Sr0.5CO3. For LiCl molten salts, strontium was carb...

Co-precipitation experiments of lanthanides were carried out using this lab-scale apparatus (4kg-salt/batch). As lanthanides, 8 lanthanide elements (Y, La, Ce, Pr, Nd, Sm, Eu and Gd) were used. By reaction with oxygen, these 8 lanthanide chlorides were converted their oxide (REO2 , RE2 O3 ) or oxychloride form. Since these lanthanide oxides or oxychlorides are nearly molten salt insoluble, they all were precipitated by free settling in the bottom of molten salt bed, where about 7–8 hrs precipitation time was requested. It was found that in the conditions of 700 °C - 12 hours sparging time and 5 L/min, all the used lanthanide elements showed over 99.5% oxidation efficiency. But in case of 800 °C molten salt temperature only after 7 hours they showed over 99% oxidation efficiency.Copyright © 2009 by ASME

Separation of some alkaline-earth chlorides (Sr, Ba) was investigated by using carbonate injection method in LiCl-KCl eutectic and LiCl molten salts. The effects of the injected molar ratio of carbonate ([K2 (or Li2 )CO3 /Sr (or Ba)Cl2 ]) and the temperature (450–750 °C) on the conversion ratio of the Sr or Ba carbonate were determined. In addition, the form of the Sr and Ba carbonate resulting from the carbonation reaction with carbonates was identified via XRD and SEM-EDS analysis. In these experiments, the carbonate injection method can remove Sr and Ba chlorides effectively over 99% in both LiCl-KCl eutectic and LiCl molten salt conditions. When Sr and Ba were co-presented in the eutectic molten salt, they were carbonated in a form of Ba0.5 Sr0.3 CO3 . And when Sr was present in LiCl molten salt, it was carbonated in the form of SrCO3 . Carbonation ratio increased with a decreasing temperature and it was more favorable in the case of a K2 CO3 injection than that of Li2 CO3 . Based on this experiment, it is postulated that carbonate precipitation method has the potential for removing alkali-earth chlorides from LiCl-KCl eutectic and LiCl molten salts.Copyright © 2007 by ASME