Hwan-Seo Park

Separation of Pure LiCl-KCl Eutectic Salt from a Mixture of LiCl-KCl Eutectic Salt and Rare-Earth Precipitates by Vacuum Distillation

In this study, the vacuum distillation of LiCl-KCl eutectic salt in a mixture of LiCl-KCl eutectic salt and rare-earth precipitates was carried out to evaluate the vaporization characteristics of LiCl-KCl eutectic salt. It was confirmed that the required time for salt vaporization was reduced by a reduction in the pressure. It appeared that the vaporization of LiCl-KCl eutectic salt containing rare-earth precipitates was decreased in comparison with that of pure salt because the salt adhered to the fine particles of the rare-earth precipitates. However, the distillation of the salt was almost achieved by elevating the surface area and further reducing the pressure. The distilled salt from the mixture consisted of 43.7 wt% LiCl and 56.3 wt% KCl. It is thought that the recovered salt can be reused because its composition is similar to the mixed ratio (44.2 wt% LiCl: 55.8 wt% KCl) of the salt used in an electrorefining process.

Erratum to: Effect of burn-up on the radioactivation behavior of cladding hull materials studied using the ORIGEN-S code

The effect of fuel burn-up on the radioactivation behavior of cladding hull materials was investigated using the ORIGEN-S code for various materials of Zircaloy-4, Zirlo, HANA-4, and HANA-6 and for various fuel burn-ups of 30, 45, 60, and 75 GWD/MTU. The Zircaloy-4 material is the only one that does not contain Nb as an alloy constituent, and it was revealed that 125Sb, 125mTe, and 55Fe are the major sources of radioactivity. On the other hand, 93mNb was identified as the most radioactive nuclide for the other materials although minor radioactive nuclides varied owing to their different initial constituents. The radioactivity of 94Nb was of particular focus owing to its acceptance limit against a Korean intermediate-/low-level waste repository. The radioactivation calculation results revealed that only Zircaloy-4 is acceptable for the Korean repository, while the other materials required at least 4,900 of Nb decontamination factor owing to the high radioactivity of 94Nb regardless of the fuel burn-up. A discussion was also made on the feasibility of Zr recovery methods (chlorination and electrorefining) for selective recovery of Zr so that it can be disposed of in the Korean repository.

Study on the phosphate reaction characteristics of lanthanide chlorides in molten salt with operating conditions

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 the case of using Li3PO4 at 550 °C. The lanthanide precipitates formed by a reaction with phosphate were a mixture of monoclinic structures, usually representing a polyhedron structure, and a tetragonal structure, representing a platelet structure.

EUTECTIC(LiCl-KCl) WASTE SALT TREATMENT BY SEQUENCIAL SEPARATION PROCESS

The sequential separation process, composed of an oxygen sparging process for separating lanthanides and a zone freezing process for separating Group I and II fission products, was evaluated and tested with a surrogate eutectic waste salt generated from pyroprocessing of used metal nuclear fuel. During the oxygen sparging process, the used lanthanide chlorides (Y, Ce, Pr and Nd) were converted into their sat-insoluble precipitates, over 99.5% at 800 °C; however, Group I (Cs) and II (Sr) chlorides were not converted but remained within the eutectic salt bed. In the next process, zone freezing, both precipitation of lanthanide precipitates and concentration of Group I/II elements were preformed. The separation efficiency of Cs and Sr increased with a decrease in the crucible moving speed, and there was little effect of crucible moving speed on the separation efficiency of Cs and Sr in the range of a 3.7 – 4.8 mm/hr. When assuming a 60% eutectic salt reuse rate, over 90% separation efficiency of Cs and Sr is possible, but when increasing the eutectic salt reuse rate to 80%, a separation efficiency of about 82 – 86 % for Cs and Sr was estimated.

A new route to the stable capture and final immobilization of radioactive cesium

Radioactive Cs released from damaged fuel materials in the event of nuclear accidents must be controlled to prevent the spreading of hazardous Cs into the environment. This study describes a simple and novel process to safely manage Cs gas by capturing it within ceramic filters and converting it into monolithic waste forms. The results of Cs trapping tests showed that CsAlSiO4 was a reaction product of gas-solid reactions between Cs gas and our ceramic filters. Monolithic waste forms were readily prepared from the Cs-trapping filters by the addition of a glass frit followed by thermal treatment at 1000°C for 3h. Major findings revealed that the Cs-trapping filters could be added up to 50wt% to form durable monoliths. In 30-50wt% of waste fraction, CsAlSiO4 was completely converted to pollucite (CsAlSi2O6), which is a potential phase for radioactive Cs due to its excellent thermal and chemical stability. A static leaching test for 28 d confirmed the excellent chemical resistance of the pollucite structure, with a Cs leaching rate as low as 7.21×10-5gm-2/d. This simple scheme of waste processing promises a new route for radioactive Cs immobilization by synthesizing pollucite-based monoliths.

Solidification of Ln Oxides Containing Volatile Chlorides from Pyrochemical Process

Abstract The waste generated from a pyrochemical process to recover uranium and transuranic elements has been one of the problematic wastes because of high volatility and low compatibility with silicate glass. For the minimization of final waste, an oxidative precipitation by sparging oxygen has been under development, and the waste containing rare earth oxides (REOs) and volatile salt is expected to be generated. This study intended to find a way to immobilize these kinds of wastes under the limitations of a processing temperature (~1200°C) and a waste loading (~20 wt%). From a series of consolidation experiments, it was induced that Ca-rich silicate glass is effective in consolidating the REOs at relatively low temperature. Based on this result, CaO-SiO2-P2O5 (CaPS) was designed to provide a way to control the volatility of waste and to avoid glass effects in the consolidation at a given temperature. By using a CaPS, REOs were consolidated, regardless of glass composition. At a high content of metal chlorides, CaPS can control the volatility up to 1200°C, but it has a low ability to immobilize alkali metal elements. For this, SiO2-Al2O3-P2O5 (SAP) was suggested to treat LiCl-KCl salt in precipitate. This composite can also control the volatility up to 1200°C, and it converted the REOs into monazite at 650°C, where the entire metal elements in chloride form are changed into relatively stable compounds. The leach test by the product consistency test-method A confirmed the immobilization ability of SAP for waste with a high content of metal chlorides. In conclusion, this study suggests the approach concept to treat a waste containing volatile compounds. For a lower content of metal chloride, CaPS are more favorable, and for a high content of metal chlorides, SAP is more effective to fabricate a wasteform for final disposal.

Al2O3-containing silver phosphate glasses as hosting matrices for radioactive iodine

ABSTRACT 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.

Separation of Lanthanide Fission Products in a Eutectic Waste Salts Delivered From Pyroprocessing of a Spent Oxide Fuel by Using Lab-Scale Oxidative Precipitation Apparatus

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

Alternative Technology for the Treatment of Waste LiCl Salt by Using Gelation with a Si-P-Al Material System and a Subsequent Thermal Conditioning Method

Abstract Chloride salt wastes, which are supposed to be generated from a pyrochemical processing of spent nuclear fuels, are one of the wastes that are problematic to treat because of their high solubility in water and the relatively high volatility of some of their nuclides during a high-temperature thermal treatment. In this paper, we propose a new conditioning method, named the gel-route stabilization/solidification (GRSS) method, and present a practical example of its application to fabricate a monolithic waste form for LiCl waste. The GRSS process is carried out in four steps: gelation, drying, mixing with binder glass, and heat treatment (thermal conditioning). The gel-forming material system consists of sodium silicate as a gelling agent, phosphoric acid as a catalyst/stabilizer, and aluminium nitrate as a promoter. Through the drying step, LiCl, CsCl, and SrCl2 are chemically converted into phosphate or aluminosilicate forms, depending on the Si/P/Al molar ratio. The gel products are thermally stable, and there is little possibility of a Cs vaporization up to 1200°C. The final waste form, fabricated by thermally treating a mixture of the gel products and borosilicate glass frit, shows low leach rates (by a product consistency test method for 7 days), 10-2 to 10-3 g/m2.day for Cs and 10-3 to 10-4 g/m2.day for Sr, which are comparable or superior to that of a glass-bonded sodalite ceramic waste form. Also, the amount of waste loading is ~16%, which is double that of the zeolite process, to generate a lesser final waste volume for disposal. From these results, it could be concluded that the GRSS method can be considered as an alternative technology for a sound immobilization of chloride salt wastes.

De-chlorination and solidification of radioactive LiCl waste salt by using SiO2-Al2O3-P2O5 (SAP) inorganic composite including B2O3 component

SAP (SiO2-Al2O3-P2O5) composite has been recently studied in KAERI to deal with the immobilization of radioactive salt waste, one of the most problematic wastes in the pyro-chemical process. Highly unstable salt waste was successfully converted into stable compounds by the dechlorination process with SAPs, and then a durable waste form with a high waste loading was produced when adding glassy materials to dechlorination product. In the present study, U-SAP composite which is SAP bearing glassy component (Boron) was synthesized to remove the adding and mixing steps of glassy materials for a monolithic wasteform. With U-SAPs prepared by a sol-gel process, a series of wasteforms were fabricated to identify a proper reaction condition. Physical and chemical properties of dechlorination products and U-SAP wasteforms were characterized by XRD, DSC, SEM, TGA and PCT-A. A U-SAP wasteform showed suitable properties as a radioactive wasteform such as dense surface morphology, high waste loading, and high durability at the optimized U-SAP/salt ratio 2.