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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.

This study is the first to report the chemical adsorption of Cs from CsI by using aluminosilicate filters developed in our laboratory. To examine the influence of oxygen on the Cs-trapping efficiency, CsI was volatilized and carried by mixed gases composed of 4% H 2 /Ar base gas and different concentrations of oxygen. Cesium capture was dramatically improved by the conversion of CsI to CsOH, which more readily formCsAlSiO 4 , in the presence of water vapor that forms if small amounts of oxygen are added to the 4% H 2 /Ar base gas.This proposed reaction mechanism was supported by thermodynamic calculations.

A reduction process in the head-end for pyroprocessing has been adopted to avoid oxidation attack on the molybdenum crucible during sintering. The reduction process is employed to reduce U3O8 pellets to UO2 prior to sintering. This allows elimination of the oxygen source, which causes oxidation attack during sintering, thereby permitting the use of a metallic crucible. However, little densification occurs due to the low reduction temperature limited by the INCONEL crucible. Consequently, the amount of material scraps from the pellets increases, thus creating an additional processing burden due to its high radioactivity. To reduce the amount of scraps, densification should be enhanced. This study suggests a simple atmospheric control strategy and clarifies its effects. With the atmospheric control, a higher bulk density and better attrition resistance were obtained in comparison to without this strategy. This can be explained in terms of O/U ratio dependent diffusion kinetics during the reduction of U3O8 to UO2.

Performance tests of mechanical decladding technology for estimating the feeding portions of the recovered fuel fragments to an electrolytic reduction process were conducted in terms of the fuel ro...

KAERI의 PRIDE 시설에서 공학규모의 전해환원용 원료물질인 UO2 다공성펠렛 제조를 위해 공정과 장치를 최적화시킨 내 용을 다루었다. UO2 분말과 별도로 attrition 밀링된 대용산화물 분말을 출발분말로, 정밀 칭량을 통해 사용후핵연료 조성을 모사하였다(Simfuel). Simfuel 분말은 각각 tumbling mixer로 혼합하여 균질화 하고, rotary press로 성형하여 furnace를 이 용해 소결하였다. 4% H2-Ar 분위기에서 1450℃ 24시간 고온 열처리하여 제조된 소결펠렛은 6.89 g·cm-3의 벌크밀도를 가 지며 이는 후속 전해환원 공정의 요구에 부합한다. 소결된 다공성펠렛의 미세구조 관찰을 통해 다공성 기지상과 함께 산화/ 금속 석출물이 관찰되어 사용후핵연료의 상이 모사됨을 확인하였다. 본 결과는 향후 공학규모 이상의 파이로 연구를 위한 UO2 다공성펠렛 제조에 중요한 기초자료로 활용 될 것이다.

A facile and simple process is described for the long-term management of Cs trapping fly-ash based filters. Glass-ceramic waste forms were synthesized from Cs-trapping filters simulated by the reaction of fly ash and Cs at a high temperature. Structural and morphological characterizations revealed that Cs was strongly immobilized in pollucite crystals, as well as in the binding glassy matrices that accommodate pollucite particles. The leaching results obtained by a product consistency test (PCT) suggested that borosilicate glass would be better than soda-lime glass to immobilize Cs-containing waste. These findings will be the basis of the final waste forms of Cs captured in fly ash filters.

A simple way to use a commercial Al 2 O 3 -based refractory in a sintering furnace was developed for the head-end system in pyroprocessing. The key was to suppress the intrinsic chemical reactivity of the mullite phase in a refractory with gaseous Cs released from spent fuel in a sintering furnace. Owing to the excellent thermodynamic and thermal stability of Y 2 O 3 , it was deposited on commercial refractories as air-plasma-sprayed coatings. The coated samples maintained their physical integrity without severe delamination after repeated heat treatment at 1450 °C for 12 h in an atmosphere of Ar and 4% H 2 . The chemical reactivity of the coated samples was also confirmed to be satisfactory through a reaction experiment using Cs 2 CO 3 powder. In addition, the coated surfaces were resistant against accidental contamination owing to the formation of non-Cs-reactive yttrium silicate rather than Cs-reactive mullite.

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.

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, a new filter was developed to capture radioactive cesium vapors evaporated from spent nuclear fuels during the pyroprocessing head-end process. For the fabrication of the filter, kaolinite powder was used as a raw material to form a fixed composition of silicon and aluminum oxide. The feasibility of the kaolinite-based filter was investigated using a system that enables the capture of gaseous cesium with respect to the trapping temperature. Characterization of the filters after cesium trapping showed that cesium vapors were chemically adsorbed in the filters by forming thermally stable cesium compounds. In addition, most of the volatilized cesium was successfully separated by the kaolinite-based filter.

We are developing a practical-scale mechanical decladder that can slit nuclear spent fuel rod-cuts (hulls + pellets) on the order of several tens of kgf of heavy metal/batch to supply UO2 pellets to a voloxidation process. The mechanical decladder is used for separating and recovering nuclear fuel material from the cladding tube by horizontally slitting the cladding tube of a fuel rod. The Korea Atomic Energy Research Institute (KAERI) is improving the performance of the mechanical decladder to increase the recovery rate of pellets from spent fuel rods. However, because actual nuclear spent fuel is dangerously toxic, we need to develop simulated spent fuel rods for continuous experiments with mechanical decladders. We describe procedures to develop both simulated cladding tubes and simulated fuel rod (with physical properties similar to those of spent nuclear fuel). Performance tests were carried out to evaluate the decladding ability of the mechanical decladder using two types of simulated fuel (simulated tube + brass pellets and zircaloy-4 tube + simulated ceramic fuel rod). The simulated tube was developed for analyzing the slitting characteristics of the cross section of the spent fuel cladding tube. Simulated ceramic fuel rod (with mechanical properties similar to the pellets of actual PWR spent fuel) was produced to ensure that the mechanical decladder could slit real PWR spent fuel. We used castable powder pellets that simulate the compressive stress of the real spent UO2 pellet. The production criteria for simulated pellets with compressive stresses similar to those of actual spent fuel were determined, and the castables were inserted into zircaloy-4 tubes and sintered to produce the simulated fuel rod. To investigate the slitting characteristics of the simulated ceramic fuel rod, a verification experiment was performed using a mechanical decladder.

A voloxidizer with a double reactor capable of processing several tens of kilograms of HM/batch of nuclear spent fuel has been developed for the decladding and voloxidation of rod-cuts into hulls and pellets through the conversion of UO2 pellets to U3O8 powder. In this study, we optimized the engineering design of this voloxidizer to improve its hull-recovery ratio. First, we tested the oxidation performance of the device prototype and evaluated the effectiveness of various mechanical and chemical voloxidizing methods. On the basis of the results, we selected the screw-and-rotation method for the double rotary drum. Next, we derived a theoretical equation for calculating the optimal reactor volume for various rod-cut weights and lengths and then validated the equation using centimeter-scale acryl reactors and hulls. Subsequently, we modularized the main components such as the heater, utility, motor, reactor, valve, and structure. The double reactor was subject to preliminary separation tests of hulls and powder. Moreover, the hull-separation performance of the voloxidizer reactor was tested at a loading of 50 kg HM/batch. Finally, the remote assembling and disassembling possibility of the modules were experimentally optimized.

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.

A practical scale mechanical decladder that can slit spent nuclear fuel rod-cuts (hulls + pellets) of several tens of kg HM/batch is being developed to supply UO2 pellets to a voloxidation process. The mechanical decladder is an apparatus for separating and recovering fuel material and cladding tubes by horizontally slitting the cladding tube of a fuel rod and a defective irradiated fuel rod. In this study, we address the engineering design of the mechanical decladder for the pretesting of rod-cut slitting. To obtain the requirements of the mechanical decladder, we first manufactured a slitter for testing based on the decladding and shearing conditions of hulls and pellets. The performance test of the testing device for decladding was carried out using a 2-CUT blade module and a 3-CUT blade module. We evaluated the decladding methods for the mechanical decladder and selected the 3-CUT blade module based on the results. A buckling measurement instrument was used to perform a buckling verification test according to the length of a rod-cut and to determine decladder dimensions. The optimum decladding rod-cut length for buckling prevention was calculated. Furthermore, we analyzed the decladding mechanism for various slitting methods. Design/fabrication and preliminary tests of the practical scale mechanical decladder were also performed. For this purpose, we constructed the main mechanism by utilizing the SolidWorks modeling and analysis program and fabricated a new mechanical decladder. Based on the derived requirements, a mechanical decladder with three main modules was designed and fabricated for testing. Simulated rod-cuts of zircaloy were also manufactured to test the basic performance of the decladder, and a data acquisition system was constructed using RSC 232 to measure decladding force and velocity. In the basic test, the rod-cut was completely sectioned into three evenly spaced locations by the new mechanical decladder.

The effect of sintering aids on the grain growth of simulated DUPIC fuel pellets with burn-ups was investigated using SIMFUEL pellets as a surrogate for actual spent PWR fuel avoiding the high radioactivity of spent fuel. In this study, titania and niobia were used as sintering aids, and SIMFUEL pellets with burn-ups of 35 GWd/tU and 60 GWd/tU were used. The sintering aids significantly improved the grain sizes of the sintered pellets. Moreover, metallic precipitates, which inhibit the movement of the grain boundary, were included even at the interior of grains due to an increase of grain boundary mobility caused by the addition of sintering aids. Titania is more effective than niobia in grain growth, while niobia is more effective than titania in densification.

Highly porous carbon microsheets (CMS) were prepared by the bulk carbonization and activation of waste packing peanuts, and applied to CO2 capture and release. Two different activation methods, physical activation using CO2 and chemical activation using KOH, were used for the preparation of porous CMS, and the effects of different activation methods were compared. Chemical activation functioned better than the physical activation, successfully producing highly microporous carbon structures with higher carbon yield. Textural properties such as specific surface area and total pore volume increased proportionally to the mass ratio of KOH/carbon because of the development of the porous structure and enlargement of pore size through KOH activation. Among the samples prepared, the highest CO2 adsorption capacities of 6.51 mol kg−1 at 0 °C and 4.07 mol kg−1 at 25 °C were obtained at 1 bar in the packing-peanut-derived CMS prepared with a KOH/carbon ratio of 3 (CMS-K3). It is noteworthy that the CO2 adsorption uptake was significantly dependent on the volume of narrow micropores of pore sizes less than 0.8 nm rather than the specific surface area or total pore volume. CMS-K3 also exhibited excellent cyclic stability, good CO2/N2 selectivity, fast adsorption kinetics, and low heat of adsorption, being regarded as a promising adsorbent for CO2 capture.

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.

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.

Recycling of LiCl–KCl eutectic salt wastes containing radioactive rare earth oxychlorides or oxides was studied to recover renewable salts from the salt wastes and to minimize the radioactive wastes by using a vacuum distillation method. Vaporization of the LiCl–KCl eutectic salt was effective above 900 °C and at 5 Torr. The condensations of the vaporized salt were largely dependent on temperature gradient. Based on these results, a recycling system of the salt wastes as a closed loop type was developed to obtain a high efficiency of the salt recovery condition. In this system, it was confirmed that renewable salt was recovered at more than 99 wt.% from the salt wastes, and the changes in temperature and pressure in the system could be utilized to understand the present condition of the system operation.

In the pyroprocess integrated inactive demonstration (PRIDE) facility at the Korea Atomic Energy Research Institute (KAERI), UO2porous pellets were fabricated as a feed material for electrolytic reduction on an engineering scale of 30 kg-U/batch. To increase the batch size, we designed and modified the corresponding equipment for unit processes based on ceramic processing. In the course of pellet fabrication, the correlation between the green density and sintered density was investigated within a compaction pressure range of 106–206 MPa, in terms of the optimization of processing parameters. Analysis of the microstructures of the produced UO2porous pellets suggested that the pellets were suitable for feed material in the subsequent electrolytic reduction process in pyroprocessing. This research puts forth modifications to the process and equipment to allow the safe mass production of UO2porous pellets; we believe these results will have immense practical interest.