In this study, neodymium (Nd) nuclides in LiCl-KC1 eutectic salts were captured and solidified using a synthetic inorganic composite (Li2O-SiO2-Al2O3-B2O3), a process that allows the selective capture of Nd and fabrication of a composite with Nd captured from waste, without additional additives or mixing. The Nd nuclides in the LiCl-KCl eutectic salt were mainly captured in the form of LiNdSiO4, and it was confirmed that NdSiOa can be formed in the composite with captured Nd when the content of Nd in the composite is increased. The capture efficiency was higher than about 98 wt%. It was thought that the salt recovered from the Nd capture test was a renewable form could be reused in the pyroprocessing of used nuclear fuel, because the composite has high chemical durability in a LiCl-KCl eutectic salt at 900 degrees C. The composite captured Nd was fabricated into a homogeneous glass form and a stable ceramic form.

The Tritium Storage & Delivery System (SDS) is part of a tokamak-type nuclear fusion reactor fuel cycle. For the safety of this cycle, the hydrogen isotopes are stored in a metal hydride form in the SDS. Depleted uranium (DU) was chosen as the storage material. DU hydride can be heated to very high temperatures that are sufficient for pumping hydrogen isotopes without using gas pumps. The experimental apparatus used to test the experimental DU bed consists of a tank that stores and measures the hydrogen, and a DU bed used for the hydriding and dehydriding of hydrogen. The DU bed is a vertical doublecylinder type with sintered metal filters. The bed is composed of primary and secondary vessels. The primary vessel contains DU, and a vacuum layer is formed between the primary and secondary vessels. In this study, recent experimental results on the pretreatment (activation and powderization) of DU and the direct hydrogen recovery and delivery of a DU bed are presented. In addition, the relationship between hydrogen pressure and temperature in the DU bed is obtained.

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.

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 UO 2 pellets to U 3 O 8 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.

The influence of cathode materials on the U electrorefining process is examined using electrochemical measurements and SEM-EDX observations. Stainless steel (STS), Mo, and W electrodes exhibit similar U reduction/oxidation behavior in 500 degrees C LiCl-KCl-UC13 molten salts, as revealed by the cyclic voltammo-grams. However, slight shifts are observed in the cathodic and anodic peak potentials at the STS electrode, which are related to the fast reduction/oxidation kinetics associated with this electrode. The U deposits on the Mo and W electrodes consist of uniform dendritic chains of U in rhomboidal-shaped crystals, whereas several U dendrites protruding from the surface are observed for the STS electrode. EDX mapping of the electrode surfaces reveals that simple scraping of the U dendrites from W electrodes pretreated in dilute HCI-l solutions to dissolve the residual salt, results in clear removal of the U deposits, whereas a thick U deposit layer strongly adheres to the STS electrode surface even after treatment. This result is expected to contribute to the development of an effective and continuous U recovery process using electrorefining. (C) 2017 Elsevier B.V. All rights reserved.

Korea is operating 24 nuclear power plants and a highly advanced neutron application reactor HANARO (High-flux Advanced Neutron Application Reactor). In addition, Korea is designing a tritium storage and delivery system (SDS) for ITER. We have been developing detritiation and tritium storage technologies since the operation of Wolsong CANDU (Canada Deuterium-Uranium) station in 1983. The Wolsong Tritium Removal System (TRF) was designed to remove tritium generated in heavy water of the moderator and heat transport. Catalysts transfer tritium from the tritiated heavy water to gaseous tritiated deuterium. The hydrogen isotopes, including tritium, are transported to a cryogenic distillation system where the tritium is removed for safe storage. Conventional high-pressure storage tanks can be dangerous for the storage of radioactive tritium gas. We have been studying various kinds of metal hydride, such as titanium, zirconium cobalt, and depleted uranium. Titanium was proven to store tritium safely and efficiently for a long period of time. Zirconium cobalt, meanwhile, incorporates tritium safely and compactly, and temporarily holds large quantities that can be recovered easily under safe, controlled conditions. However owing to the disproportionation characteristics of zirconium cobalt, we are now developing depleted uranium hydride safe handling technologies. In this technical note, we present the details of the recent development progress of these tritium systems.

사용후핵연료 파이로프로세싱에서는 방사성 희토류 염화물(RECl3)을 함유한 LiCl-KCl 공융염폐기물이 발생되며, 핫셀시설 에서 운영을 목적으로 단순한 형태의 공융염폐기물 처리공정을 개발하는 것이 필요하다. 본 연구에서는, LiCl-KCl 공융염폐 기물 내 희토류 핵종 분리/고화공정의 단순화를 목적으로 Li2O-Al2O3-SiO2-B2O3계의 무기합성매질을 이용하여 LiCl-KCl 공융 염 내 희토류 핵종(Nd)을 분리한 후 분리생성물을 바로 고화하는 시험을 실시하였다. 공융염 내 희토류 염화물(NdCl3) 대비 0.67의 무게비에 해당하는 무기합성매질의 양으로도 Nd 핵종을 98wt% 이상 분리할 수 있었고, 이 때 얻은 희토류 핵종 포 집생성물은 약 50wt% 수준의 희토류 산화물 함량을 보유하고 있었으며, 이 포집생성물을 화학적 내구성이 우수한 단일상의 균질한 유리고화체로 제조할 수 있었다. 이 결과들은 LiCl-KCl 공융염폐기물 내 희토류 핵종의 분리/고화공정을 단순화하기 위한 방안수립에 활용될 수 있을 것이다.

Estimation of centerline temperature of nuclear glass waste form for each waste stream is very essential in the period of storage because the centerline temperature being over its glass transition temperature results in the increase of leaching rate of radioactive nuclides due to the devitrification of glass waste form. Here, to verify the effects of waste form diameter and transuranic element content in the rare earth waste on the centerline temperature of the waste form, the surrogate rare earth glass waste generated from pyrochemical process was immobilized with SiO2-Al2O3-B2O3 glass flit system, and thermal properties of the rare earth glass waste form were determined by thermomechanical analysis and thermal conductivity analysis. The estimation of centerline temperature was carried out using the experimental thermal data and steady-state conduction equation in a long and solid cylinder type waste form. It was revealed that thermal stability of waste form in case of 0.3 m diameter was not affected by the TRU content even in the case of 80% TRU recovery ratio in the electrowinning process, meaning that the waste form of 0.3 m diameter is thermally stable due to the low centerline temperature relative to its glass transition temperature of the rare earth glass waste form. (C) 2016 Elsevier B.V. All rights reserved.

The pyrochemical process for the recycling of used nuclear fuels generates LiCl and LiCl–KCl waste salt during a series of processes. In particular, for the reduction of LiCl–KCl waste volume, radioactive rare earth nuclides in the LiCl–KCl waste salt are separated by a phosphorylation precipitation reaction, generating rare earth phosphate waste in a powder form. In this study, for the immobilization of rare earth phosphate wastes, rare earth calcium phosphate (RCP) glass waste forms have been fabricated using a CaO–P2O5 glass system under a moderate vitrification temperature of below 1200°C. Vitrification properties of NdPO4, which is a representative rare earth phosphate material have been investigated through XRD, SEM, EDX, DSC, IR, and density measurements. The RCP glass waste form fabricated from the 4CaO–6P2O5 system showed an immobilization capacity of about 32wt.% NdPO4 waste loading. Also, chemical durabilities of RCP glass waste forms are investigated by PCT-A test.

The effect of fluoride compounds on Zr electrorefining is investigated for Zirlo cladding tubes in LiCl-KCl based molten salts. Cyclic voltammetric curves exhibit a modification of electrochemical behaviors of Zr ions from a two-step reduction process in LiCl-KCl-0.07 M ZrCl4 salts to one-step reduction by the addition of 0.84 M LiF. Morphological features of Zr deposits are found to also be changed from a powder type to a dendritic characteristic, thereby reducing the amount of incorporated salts and enhancing the crystallinity of electrorefined Zr. Quantitative measurement reveals that the Zr content in the deposit is increased by the addition of fluoride compounds, which is expected to enhance the recovery yield of Zr.

In this study, the effect of cold drawing and heat treatment on the microstructure of Invar36 alloy wire was investigated. Invar36 alloy wire is used as a transmission line core material, and is required to have high strength. The diameter of the Invar36 alloy wire specimens were reduced from 16 mm to 4.3 mm after three cold drawing and two heat treatment processes, thereby increasing tensile strength. Specimens were taken after each of the cold drawing and heat treatment processes, and their microstructure and tensile properties were analyzed. The Invar36 alloy wire had a gamma(-)(Fe, Ni) phase matrix before the cold drawing and heat treatment processes. After the cold drawing processes, {220} and {200} textures were mainly achieved. After the heat treatment processes, a {220} recrystallization gamma(-)(Fe, Ni) phase was formed with fine carbides. The recrystallization gamma(-)(Fe, Ni) phase grains had low dislocation density, so they probably accommodated a large amount of deformation during the cold drawing processes.