In KAERI, the nuclide management technology is currently being developed for the reduction of disposal area required for spent fuel management. Among the all fission products of interest, Cs, I, Kr, Tc are considered to be significantly removed by following mid-temperature and hightemperature treatment, however, a difficulty of real spent-fuel thermal treatment experiment limits the development of such thermal treatment. The test employing SimFuel (Simulated Spent Fuel) can be an alternative for such condition, however, the fabrication of SimFuel containing semivolatile species such as Cs, I and Re (substitute for Tc) was not achieved for conventional sintering method since such species are easily removed during hot temperature treatment. In this study, for the prevention of volatilization of such species and the inclusion of semi-volatile species in fabrication of SimFuel, argon-based high pressurizing up to Max 100 bar was considered to be applied in high temperature treatment. For this, lab-scale hot-isostatic press applicable up to 1,500°C was fabricated and is being waiting for the approval for high-pressure test. After approval of license, UO2 baesd SimFuel containing CsI will be fabricated and its micro-structure and composition will be evaluated through SEM-EDX and XRD
The stabilization techniques are highly required for damaged nuclear fuel to strengthen safety in terms of transportation, storage, and disposal. This technique includes recovering fuel materials from spent fuel, fabrication of stabilized pellets, and fabrication of fuel rods. Thus, it is important to identify the leaching behavior of the stabilized pellets to verify their stability in humid environments which are similar to storage conditions. In this study, we introduce various leaching experiment methods to evaluate the leaching behavior of the stabilized pellets, and determine the most suitable leaching test methods for the pellets. Also, we establish the leaching test conditions with various factors that can affect the dissolution and leaching behavior of the stabilized pellets. Accordingly, we prepare the simulated high- (55 GWd/tU) and low- (35 GWd/tU) burnup nuclear fuel (SIMFUEL) and pure UO2 pellets sintered at 1,550°C and 1,700°C, respectively. Each pellet is placed in a vessel and filled with DI water and perform the leaching test at three different temperature to verify the leaching mechanism at different temperature range. Based on the standard leaching test method (ASTM C1308-21), the test solution is removed from the pellet after specific time intervals and replaced in the fresh water, and the vessel is placed back into the controlled-temperature ovens. The test solutions are analyzed by using ICP-MS.
The damaged spent fuel rods must be stabilized by encapsulation or dry re-fabrication technologies before geological disposal. For applying the dry re-fabrication technology, we manufactured a vertical type furnace to perform the fuel material recovery from damaged fuel rods by oxidative decladding technology. As driving forces to accelerate oxidative decladding rate, magnetic vibration and pulse hammering generated by a pneumatic cylinder were used in this study. The oxidative decladding efficiency and recovery rate of fuel oxide powder with rod-cut length, oxidation temperature and time, oxygen concentration, and gas mixtures were investigated using simfuel rod-cuts in a vertical furnace for fuel material recovery and powder quality improvement. The oxidative decladding was performed for 2.5-10 h as following operation parameters: simfuel rod-cut length of 50-200 mm, oxidative temperature from 450 to 580°C, oxygen concentration of 49.5 or 75.6%, and gas mixtures in O2/Ar or O2/N2. In magnetic vibration, oxidative decladding was progressed only at bottom portion of fuel rodcut. Whereas, oxidative decladding in pulse hammering was occurred at both top and bottom portions of fuel-rod. In pulse hammering method, the oxidative decladding conditions to declad rod-cuts of 50- 200 mm in length were established to achieve both decladding efficiency of ~100% and fuel material recovery rate of > 99%. These conditions were as follows: oxidation temperature and time at 500°C and 2.5-10 h, oxygen concentration at 75.6% under O2/N2 gas mixtures. As operation conditions for a pneumatic cylinder, stroking, actuating, and waiting times were 0.5, 3, and 12 s.
Irradiated uranium dioxide in damaged used fuel could oxidize during transportation, interim storage or disposal, resulting that the fuel pellet fragments are reduced to a grain-sized powder that can easily escaped from the damaged rod. It has been reported that oxidized spent fuel (i.e. U4O9+x) that was in contact with water could increase the dissolution rate by making the grain boundaries more accessible to the water. Therefore, the damaged used fuel requires stabilization technology including nuclear material recovery, pellet manufacturing process, and stabilization fuel rod manufacturing that can secure safety in terms of permanent disposal. In this study, we prepared pure UO2 and SIMFUEL pellets that are a mixture of UO2 and surrogated metallic oxides for fission products equivalent to a burn-up of 35 GWd/tU and 55 GWd/tU as the stabilized spent fuel. The UO2 and fission products powders were milled and pressed into pellets at 250 MPa and sintered at 1,550°C and 1,700°C for 6 hours in an atmosphere of 4%H2-Ar. The prepared UO2 and SIMFUEL pellets were placed in PTFE Teflon vessels and filled with deionized water to identify the leaching behavior by a long-term leaching experiment under the similar condition to a repository for the safe disposal.
To estimate the removal efficiency of TRU and rare earth elements in an oxide spent fuel, basic dissolution experiments were performed for the reaction of rare earth elements from the prepared simfuel with chlorination reagents in LiCl-KCl molten salt. Based on the literature survey, NH4Cl, UCl3, and ZrCl4 were selected as chlorination reagent. CeO2 and Gd2O3 powders were mixed with uranium oxide as a representative material of rare earth elements. Simfuel pellets were prepared through molding and sintering processes, and mechanically pulverized to a powder form. The experiments for the reaction of the simfuel powder and chlorination reagents were carried out in a LiCl-KCl molten salt at 500°C. To observe the dissolution behavior of rare earth elements, molten salt samples were collected before and after the reactions, and concentration analysis was performed using ICP. After the reaction completed, the remaining oxide was washed with water and separated from the molten salt, and XRD was used for structural analysis. As a result of salt concentration analysis, the dissolution performance of rare earth elements was confirmed in the reaction experiments of all chlorination reagents. In an experiment using NH4Cl and ZrCl4, the uranium concentration in the molten salt was also measured. In other words, it seemed that not only rare elements but also uranium oxide, which is a main component of simfuel, was dissolved. Therefore, it is thought that the dissolution of rare earth elements is also possible due to the collapse of the uranium oxide structure of the solid powder and the reaction with the oxide of rare earth elements exposed to molten salt. As a result of analyzing the concentration changes of Simfuel before and after each reaction, there was little loss of uranium and rare earth elements (Ce/Gd) in the NH4Cl experiment, but a significant amount of rare earth elements were found to be reduced in the UCl3 experiment, and a large amount of rare earth elements were reduced in the ZrCl4 reaction.
최대 선출력 61 ㎾/m 및 평균 연소도 1,770 ㎿d/tU의 조건으로 하나로에서 조사한 DUPIC(Direct Use of Spent PWR Fuel in CANDU Reactors) 핵 연료를 EPMA (Electron Probe Micro Analyzer)를 이용하여 핵분열 생성물을 분석하였다. EPMA의 정확한 분석 방법을 확립하고자, 핵분열생성물 대신 시약을 첨가하여 제조한 모의 DUPIC 핵연료로 EPMA 분석을 수행하였고, 그 결과를 습식 화학 분석의 결과와도 비교하여 평가하였다. 모의 DUPIC 핵연료 중심부의 금속 석출물은 약 1 정도의 크기로 관찰되었으며, 이들의 조성은 Mo-53.89 at.%, Ru-37.40 at.% 및 Pd+Rh-8.71 at%이었다. 모의 DUPIC 핵연료 시험에서 정립한 시험방법으로 조사한 DUPIC 핵연료 시편의 금속 석출물 특성을 분석하였다. 핵연료 중앙부에서 관찰된 금속 석출물들의 크기는 2∼2.5 정도이었으며, Mo-47.34 at.%, Ru-46 at.%, Pd+Rh-6.65 at.%의 조성임을 확인하였다. 이 실험을 위하여, 특별히 시료의 전도성을 향상시키기 위한 처리를 하였으며, 작은 금속 석출물에 EPMA의 전자빔을 정확히 조사할 수 있는 실험 조건을 제시하였다.