Vitrification is one of the best ways to immobilize high-level radioactive waste (HLW) worldwide over the past 50 years. Since the glass matrix has a medium (3.0-5.5 A) and short (1.5-3.0 A) periodicity, it can accommodate most elements from the periodic table. Borosilicate glass is the most suitable glass matrix for vitrification due to its high chemical durability, high waste-loading capacity, and good radiation resistance. Mo is a fission product contained in liquid waste generated in the process of reprocessing spent nuclear fuel and exists in the form of MoO4 2- in the glass. MoO4 2- forms a depolymerization region without directly connecting with the glass network former. When the concentration of Mo increases in the depolymerization region, it combines with nearby alkali or alkaline earth cations to form a crystalline molybdate phase. Phase separation and crystallization in the glass can degrade the performance of the material because it changes the physical and chemical properties of the glass. In particular, since alkali molybdate has high water solubility when it forms crystals containing radioactive elements such as Cs, there is a risk of leakage of radionuclides by groundwater during deep underground disposal. Therefore, in this study, the most stable glass-ceramic composition was developed using various alkali elements, and the difference in phase separation and crystallization behavior in glass and the stability of the solidified body were analyzed by structural analysis of the glass network and alkali molybdate. The cause of the difference in crystallization of alkali molybdate according to the type of alkali cation is structurally analyzed, and using this, research is conducted to increase the Mo content in the glass without crystallization.

The stabilization technology for the damaged spent fuel is being developed to process the damaged fuel into sound pellet suitable for dry re-fabrication. It requires several treatments including oxidative decladding followed by reduction treatment for oxidized powder closely related to the quality of oxidized powders for pellet fabrication. For the development of operating condition for the reduction treatment, in this study, we evaluated the effect of air-cylinder based vertical shaking previously applied to oxidative decladding on powder reduction. For U3O8 of 50-100 g, the reduction test were applied with and without vertical shaking at 700°C under reduction atmosphere (Ar + 4%H2) and the concentration of hydrogen in effluent was measured to evaluate the reduction reaction. It was found that the vertical shaking system has allowed the reaction time of 50 g and 100 g U3O8 reduced by 33% compared to the test in static mode. Based on XRD analysis, the better crystallinity of the products was also achieved.

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.

Background: In order to improve the daily life of stroke patients, it is necessary to restore walking performance. Improving gait performance is one of the main goals for stroke rehabilitation. Objectives: To investigate the change in gait parameters through evaluation of one stroke patient before, during, and after wearing the weight vest. Design: A case study. Methods: This study is an ABA study design. In principle, one assessment per day (A1) was measured three times without wearing the weight vest. Intervention assessment (B) was performed 6 times while wearing the weight vest. The second baseline evaluation (A2) was measured three times with the weight vest removed again. Results: The gait speed was increased when the weight vest was removed than when the weight vest was worn. Also, gait symmetry increased when wearing the weight vest. Conclusion: Wearing the weight vest can reduce the gait asymmetry in stroke patients and increase the gait speed after wearing the weight vest.

수십 년간의 세계화에도 불구하고, 기업들은 여전히 국내와 해외 시장에서의 마케팅 환경의 불일 치로 어려움을 겪고 있다. 기업들은 이러한 환경의 차이를 극복하기 위해 각 시장의 특성을 고려한 광고 어필(advertising appeal)을 이용하여 특정 메시지를 전달한다. 이에 따라 기업의 국내와 해외 시장에서의 광고 어필 사용의 차이가 존재하는데, 그 차이를 찾기 위해 본 연구는 한국과 말 레이시아에서 한국브랜드가 어떤 광고 어필을 활용하는지 살펴보려고 한다. 이번 연구는 유튜브 동 영상의 콘텐츠 분석을 통해 한국과 말레이시아 시장 간 광고 어필 사용의 차이를 도출하였다. 본 논문은 말레이시아 시장에서의 마케팅 커뮤니케이션 전략을 계획할 때 국내 마케팅 매니저들이 고 려할 수 있는 실질적인 시사점을 제시한다.

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 high-temperature treatment, however, a difficulty of spent-fuel thermal treatment experiment limits the development of such thermal treatment. In this study, we applied our previously developed two-stage diffusion release model coupled to UO2 oxidation model to the development of optima thermal treatment scenario. Since the formation of cesium pertechnetate should be avoided and the fission release behavior is considerably affected by the extent of oxygen, we obtained oxygen-content dependent model parameters for two-stage fission release model and applied the model to the evaluation of fission release behavior to different oxygen content and thermal treatment procedure. It was found that the developed fission release model closely describes the experimental behavior of fission product of interest, implying a validity of model prediction and the thermal treatment condition reducing the chemical reaction between cesium and technetium could be developed.