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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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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