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.

We have developed a practical-scale dry disassembling process to dismantle PWR (Pressurized Water Reactor) spent nuclear fuel assembly in the order of several tens of kilograms of heavy metal/batch to supply rod-cuts (cladding tube and UO2 pellets) for mechanical decladding process. Dry head-end disassembling process has advantages over the wet head-end process because of the lower risk of proliferation and treatment of spent fuel with relatively high heat and radioactivity. This study describes the main design considerations for the disassembling process of the spent nuclear fuel assembly during the dry head-end process. The down-ender, dismantling, extraction, and cutting technologies are analyzed and models have been designed for testing. The purpose of dry head-end disassembly process is to test the main device performance and to obtain scale-up data for practical-scale disassembling. With this in mind, design considerations were analyzed based on remoteness, and basic verification tests were performed. However, the authors used simulated fuel, instead of the actual spent fuel, owing to a lack of joint determination. In addition, in the present study, we did not consider the heat generated from minor actinides or the radioactivity of the fission product; these aspects will be considered in a future study. During the basic test performed in this study, a simulated assembly was completely disassembled using new methods, such as dismantling, extraction, and cutting processes. The practical-scale dry disassembling technology can be tested using scale-up data for reuse of the spent fuel.

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...

Ag-containing 1 hydrophobic aluminosilicate aerogels capable of capturing I2 were prepared by sol-gelation from Na, Al, and Si alkoxides using co-solvent exchange, Ag/Na ion exchange, solvent exchange, silylation-induced surface hydrophobization, and ambient-pressure drying. The Ag contents, Brunauer-Emmett-Teller (BET) surface areas, and pore size distributions of aerogels depended on the solvent of the co-solvent exchange step, with isopropanol affording higher Ag contents and lower BET surface areas than water. This behavior was ascribed to the reaction of isopropanol-insoluble NaOH in gel pores during Ag/Na exchange with aqueous AgNO3 to form water-insoluble Ag2O. The Ag-containing hydrophobic aluminosilicate aerogel modified with hexamethyldisilazane contained meso- and macropores and exhibited a maximum iodine capture capacity of 0.555 g-I/g-sorbent, a BET surface area of 172 m2/g, a specific pore volume of 1.29 cm3/g, and an average pore size of 23.4 nm.

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.

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.

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.

The electrolytic reduction process in pyroprocessing of used fuel requires that fuel fragments be pre-treated to remove cesium and iodine. The effect of high temperatures and fuel fragment size on release of cesium was investigated over 10 h at a temperature of 1300–1400 °C for fuel fragments with an O/U ratio of 2.2. Re-fragmentation of the original fragments was observed as changes in cesium release count rates as fuel was heated from 484 to 1329 °C. The release of cesium is highly dependent on the size of fuel fragments and most of the cesium is released as the temperature is held at the maximum target value.

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.

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.

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.

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.

In this paper, an improvement study on the closed chamber distillation system for recovery of renewable salts from salt wastes containing radioactive rare earth compounds was performed to determine optimum operating conditions. It was very important to maintain the pressure in the distillation chamber below 10 Torr for a high efficiency (salt recovery >99 %) of the salt distillation. This required increasing the salt vaporization and condensation rates in the distillation system. It was confirmed that vaporization and condensation rates could be improved controlling the given temperature of top of the condensation chamber. In the distillation tests of the salt wastes containing rare earth compounds, the operation time at a given temperature was greatly reduced changing the given temperature of top of the condensation chamber from 780 to 700 °C.

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 this paper, a separation method of radionuclides (Ba, Sr) from LiCl salt wastes generated from the electroreduction process of spent nuclear fuel was studied to recover pure LiCl salts and reduce radioactive wastes. The method consisted of chemical conversion process of BaCl2 and SrCl2 in LiCl molten salts by using lithium compounds and vacuum distillation process of LiCl salts. In the chemical conversion, BaCl2 and SrCl2 in LiCl molten salts were mainly converted into (Ba,Sr)CO3 or (Ba,Sr)SO4. Contents of Ba and Sr in LiCl salts recovered from the vacuum distillation process were equal to about 0.01 of initial concentrations of Ba and Sr in LiCl molten salts. These results will be utilized to recycle the LiCl salt wastes.

Pyroprocessing technology was developed in the beginning for metal fuel treatment in the US in the 1960s. The conventional aqueous process, such as PUREX, is not appropriate for treating metal fuel. Pyroprocessing technology has advantages over the aqueous process: less proliferation risk, treatment of spent fuel with relatively high heat and radioactivity, compact equipment, etc. The addition of an oxide reduction process to the pyroprocessing metal fuel treatment enables handling of oxide spent fuel, which draws a potential option for the management of spent fuel from the PWR. In this context, KAERI has been developing pyroprocessing technology to handle the oxide spent fuel since the 1990s. This paper describes the current status of pyroprocessing technology development at KAERI from the head-end process to the waste treatment. A unit process with various scales has been tested to produce the design data associated with the scale up. A performance test of unit processes integration will be conducted at the PRIDE facility, which will be constructed by early 2012. The PRIDE facility incorporates the unit processes all together in a cell with an Ar environment. The purpose of PRIDE is to test the processes for unit process performance, operability by remote equipment, the integrity of the unit processes, process monitoring, Ar environment system operation, and safeguards related activities. The test of PRIDE will be promising for further pyroprocessing technology development.

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.

Korea is operating 20 nuclear units, producing about 700 t/yr of spent fuel. This spent fuel should be managed in a safe and practicable way. One option for this spent fuel management is treatment with pyrochemical process. Pyrochemical processflow sheetin KAERI includes voloxidation, electroreduction, electrorefining, electrowinning, and waste salt treatment processes. Each unit process has been tested in lab-scale, and these unit processes will be incorporated in PRIDE (PyRoprocess Inactive integrated DEmonstration) Facility in late 2011 in order to evaluate scaled-up design on the basis of the information of lab-scale experiments, integrity of each unit performance, and safeguards of integrated process, etc.

A fundamental study on the distillation rate on LiCl-KCl eutectic salt under different vacuums from 66 to 6600 Pa (0.5 to 50 mm Hg) was performed by using both a nonisothermal and an isothermal thermogravimetric (TG) analysis. Based on the nonisothermal TG data, distillation rate equations as a function of the temperature could be derived. Calculated flux by these model flux equations was in agreement with the distillation rate obtained from isothermal TG analysis. A salt distillation operation with a moderated distillation rate of 10- 4 to 10- 5 mol·cm -2 ·s -1 is possible at temperatures of

As an alternative to conventional Group I and II separation methods (such as adding a chemical agent and ion exchange), melt crystallization processes, zone freezing, and layer melt crystallization were tested for the separation (or concentration) of cesium and strontium fission products in a LiCl waste salt generated from an electrolytic reduction process of a spent oxide fuel. In these melt crystallization processes, impurities (CsCl and SrCl 2 ) are concentrated in a small fraction of the LiCl salt by the solubility difference between the melt phase and the crystal phase. As experimental variables, initial molten salt temperature, crucible rising velocity in the zone freezing case, and cooling air flow rate in the layer crystallization case were used. In the zone freezing process, although the operating time is long (1.7 mm/h of crucible rising velocity) when assuming a LiCl salt reuse rate of 90 wt%, >90% separation efficiency for both CsCl and SrCl 2 was shown. In the layer crystallization process, the crystal growth rate strongly affects the crystal structure and therefore the separation efficiency. At a 25 to 30 l/min cooling air flow rate, 700 to 710°C initial molten salt temperature, and

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.

Distillation and condensation characteristics of LiCl-KCl eutectic salts containing rare earth precipitates were investigated and a conversion of the rare earth oxychlorides to oxides was performed. The distillation flux was increased by about 1,000 times by reducing the pressure from 760 Torr to 0.5 Torr. The composition of the recovered salts was changed according to the condensed spot. The conversion temperature of the multicomponent oxychlorides was decreased when compared to that of a single component. It was found that the concentration of the chloride gas in the flue gas is an index to establish whether the conversion is completed or not.

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...

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.

This report summarizes treatment and waste form options being evaluated for waste streams resulting from the electrochemical/pyrometallurgical (pyro ) processing of used oxide nuclear fuel. The technologies that are described are South Korean (Republic of Korea – ROK) and United States of America (US) ‘centric’ in the approach to treating pyroprocessing wastes and are based on the decade long collaborations between US and ROK researchers. Some of the general and advanced technologies described in this report will be demonstrated during the Integrated Recycle Test (IRT) to be conducted as a part of the Joint Fuel Cycle Study (JFCS) collaboration between US Department of Energy (DOE) and ROK national laboratories. The JFCS means to specifically address and evaluated the technological, economic, and safe guard issues associated with the treatment of used nuclear fuel by pyroprocessing. The IRT will involve the processing of commercial, used oxide fuel to recover uranium and transuranics. The recovered transuranics will then be fabricated into metallic fuel and irradiated to transmutate, or burn the transuranic elements to shorter lived radionuclides. In addition, the various process streams will be evaluated and tested for fission product removal, electrolytic salt recycle, minimization of actinide loss to waste streams and wastemore » form fabrication and characterization. This report specifically addresses the production and testing of those waste forms to demonstrate their compatibility with treatment options and suitability for disposal.« less

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