• Nuclear Engineering and Technology
• /
• v.26 no.3
• /
• pp.374-380
• /
• 1994

A method to simulate the extent of interlinkage of grain boundary gas bubbles to the free surfaces of fuel pellet was developed. With the shape of UO$_2$gain treated as tetrakaidecahedron (TKD)), the interlinked fraction of fission gas bubbles to free surfaces at grain comers was calculated as a function of the radius of grain corner bubbles by the Monte Carlo technique. In spite of two dimensional analysis, the present method shooed reasonable agreement between predicted and measured fuel swelling at the moment that complete bubble interlinkage was achieved. However, for more realistic simulation of interlinkage, grain comer bubbles should be treated three dimensionally.

• Nuclear Engineering and Technology
• /
• v.31 no.6
• /
• pp.608-617
• /
• 1999

The powder mixture which has a bimodal size distribution, with a large mode corresponding to AUC-UO$_2$ powder and a small one corresponding to ADU-UO$_2$ powder, was prepared, pressed into compacts, and sintered at 1680t for 4 hours in hydrogen gas. The compact density of the powder mixture increases with increasing ADU-UO$_2$content within a content of 20 wt %, since small ADU-UO$_2$ particles can fill interstices between large AUC-UO$_2$particles. The UO$_2$ pellet made using the powder mixture has a lower open porosity than that made using AUC-UO$_2$ powder alone. The mechanism for the formation of a flake-like pore is proposed, and the decrease in open porosity may be ascribed to the decrease in the number of flake-like pores.

• Nuclear Engineering and Technology
• /
• v.50 no.2
• /
• pp.259-267
• /
• 2018

Focused ion beam-scanning electron microscope and electron backscattered diffraction examinations were conducted in the center of a $73\;GWd/t_U\;UO_2$ fuel. They showed the formation of subdomains within the initial grains. The local crystal orientations in these domains were close to that of the original grain. Most of the fission gas bubbles were located on the boundaries. Their shapes were far from spherical and far from lenticular. No interlinked bubble network was found. These observations shed light on previous unexplained observations. They plead for a revision of the classical description of fission gas release mechanisms for the center of high burn-up $UO_2$. Yet, complementary detailed observations are needed to better understand the mechanisms involved.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.15 no.4
• /
• pp.343-353
• /
• 2017

Processing and equipment were tailored for engineering scale fabrication of $UO_2$ porous pellets, a feed material for the electrolytic reduction process in the PRIDE (PyRoprocessing Integrated DEmonstration) facility at KAERI (Korea Atomic Energy Research Institute). The starting materials, $UO_2$ powder and pre-milled surrogate oxide powders, were proportioned to simulate the chemical composition of spent fuel (so-called Simfuel). The Simfuel powders were homogenized by mixing, compacted into a pellet shape, and finally heat treated using a tumbling mixer, rotary press, and sintering furnace. After sintering at $1450^{\circ}C$ for 24 h in $4%\;H_2-Ar$, the average bulk density of the $UO_2$ Simfuel pellets was $6.89g{\cdot}cm^{-3}$, which meets the standard of the following electrolytic reduction process. In addition, the results of a microstructural analysis demonstrated that the sintered Simfuel $UO_2$ porous pellets accurately simulate the properties of spent fuel in terms of the formation of second phases. These results provide essential information for the massive fabrication of $UO_2$ porous pellets for engineering scale pyroprocessing research.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.17 no.1
• /
• pp.37-46
• /
• 2019

As candidates for accident-tolerant fuels, ceramic microcell fuels, which are distinguished by their peculiar microstructures, are being developed; these fuels have $UO_2$ grains surrounded by cell walls. They contribute to nuclear fuel safety by retention of fission products within the $UO_2$ pellet, reducing rod pressure and incidence of SCC failure. Cesium, a hazardous fission product in terms of amount and radioactivity, can be captured by chemical reactions with ceramic cell materials. The capture-ability of cesium therefore depends on the thermodynamics of the capturing reaction. Conversely, compositional design of cell materials should be based on thermodynamic predictions. This study proposes thermodynamic calculations to evaluate the cesium capture-ability of three ceramic microcell compositions: Si-Ti-O, Si-Cr-O and Si-Al-O. Prior to the calculations, the chemical and physical states of the cesium and the cell materials were defined. Then, the reactivity was evaluated by calculating the cesium potential (${\Delta}G_{Cs}$) and oxygen potential (${\Delta}G_{O_2}$) under simulated LWR circumstances of normal operation. Based on the results, cesium capture is expected to be spontaneous in all cell compositions, providing a basis for the compositional design of ceramic microcell fuels as well as a facile way for evaluating cesium capture.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.16 no.3
• /
• pp.291-299
• /
• 2018

Under a pyro-processing concept, an electrolytic reduction process has been developed to reduce uranium oxide in molten salt by electrochemical means as a part of spent fuel treatment process development. Accordingly, a model based on electrochemical theory is required to design a reactor for the electrolytic reduction process. In this study, a 1D model based on the phase-field theory, which explains phase separation behaviors was developed to simulate electrolytic reduction of uranium oxide. By adopting parameters for diffusion of oxygen elements in a pellet and electrochemical reaction rate at the surface of the pellet, the model described the behavior of inward reduction well and revealed that the current depends on the internal diffusion of the oxygen element. The model for the electrolytic reduction is expected to be used to determine the optimum conditions for large scale reactor design. It is also expected that the model will be applied to simulate the integration of pyro-processing.

• Nuclear Engineering and Technology
• /
• v.28 no.4
• /
• pp.349-358
• /
• 1996

The leaching behavior of uranium from unirradiated CANDU UO$_2$ fuel pellet in the spent fuel wet storage and disposal conditions has been investigated. A modified IAEA leach test method was used, and then the extent of leaching was monitored by analysis for uranium in the leachant. The leach test has been performed in various leachants(demineralized water and boric acid solution at pH=6, synthetic granite groundwater) for a long-term period of 5.4 years, and the effect of temperature on the leach rate of uranium has been analyzed. The leach rates of uranium at $25^{\circ}C$ were dependent on the leachants. Over initial 100 days of leach periods, the leach rate in groundwater was the highest in three leachants and no significant differences of leach rates ore observed in the demineralized oater and boric acid solution. But these leach rates in three leachants around 2,000 days at $25^{\circ}C$ appeared to be reached the steady rates in the range of 1~5$\times$10$^{-8}$ g/$\textrm{cm}^2$ day. The leach rate of uranium in groundwater shooed to be independent of the temperature, but those in both demineralized water and boric acid solution increased with temperature. These results show that the leaching behavior of uranium from UO$_2$ fuel in both the demineralized water ann boric acid may be controlled tv the surface oxidative.dissolution reaction of UO$_2$ and the leach rate of uranium in groundwater at room temperature could mainly be controlled by the complex reaction of dissolved uranyl ions with carbonate ions and no variation of leach rate of UO$_2$ in groundwater with temperature may be due to the local deposition of passivating uranyl phases on the surface.

• Nuclear Engineering and Technology
• /
• v.31 no.5
• /
• pp.455-464
• /
• 1999

The effect of TiO$_2$ on the sintering behavior of mixed UO$_2$-U$_3$O$_{8}$ Powder compacts has been investigated using the U$_{3}$O$_{8}$ powder made tv oxidation of defective UO$_{2}$ pellets. Without TiO$_2$, UO$_2$ pellet density is inversely proportional to U$_3$O$_{8}$ content and is below 94 %TD in the U$_3$O$_{8}$ range above 15 wt%. Using more than 0.1 wt % TiO$_2$, however, the density decreases slightly with U$_3$O$_{8}$ content and thus is higher than about 94% TD in the whole range of U$_3$O$_{8}$ content. The grain sizes of UO$_2$ pellets with more than 0.1 wt % TiO$_2$are larger than about 30${\mu}{\textrm}{m}$. Therefore, the U$_3$O$_{8}$ Powder can be reused without any restriction on its amount in UO$_2$ pellet fabrication by sintering the mixed UO$_2$-U$_3$O$_{8}$ compact with the aid of TiO$_2$. Mechanisms for densification and grain growth are proposed and discussed, based on a dilatometry study and an examination of microstructure. microstructure.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2003.11a
• /
• pp.438-446
• /
• 2003

This paper presents the development of the remotely operated decontamination systems for use in a highly radioactive zone of the DUPIC Fuel Development facility of the Irradiated Material Examination Facility at the Korea Atomic Energy Research Institute. The remotely operated decontamination systems were designed to completely eliminate human interaction with hazardous radioactive contaminants. These decontamination systems are mainly classified into three systems depending on the task environment - a fabrication equipment decontamination system, a hot-cell floor decontamination system, and an isolation room floor decontamination system. A decontamination system for contaminated fabrication equipment utilizes dry ice pellet blasting method to decontaminate contaminated surface of the equipment. The decontamination systems for the hot-cell floor and isolation room floor employ a vacuum cleaning method to decontaminate the contaminated floor and collect loose dry spent nuclear fuel debris and other radioactive waste placed on the floor. The human operator from the out-of-cell performs a series of decontamination tasks remotely by manipulating decontamination systems located in-cell via a handcontroller with the aid of vision feedback information. The environmental, functional and mechanical design considerations, control system and capabilities of the remotely operated decontamination systems at a high radioactive environment are also described.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
• /
• 2007.05a
• /
• pp.361-365
• /
• 2007

우라늄을 내장한 연료봉은 핵분열이 일어나는 우라늄 펠렛(pellet)을 1차적으로 차폐하는 중요한 구조물이다. 연료봉은 원자로 내에서 유체유발진동에 의해 손상될 수 있으며, 본 연구에서는 유동유발진동 특성을 예측하기 위해 핵연료봉의 동특성 규명을 위한 모드해석을 수행하였다. 핵연료봉의 진동특성을 규명하기 위해 제작한 시험장치를 이용하여 피복관(clad tube)의 진동특성실험과 유한 요소 해석을 수행하였다. 모드시험(Modal Testing)은 현재 상용 핵연료봉(튜브)을 대상으로 수행되었으며, 유한 요소 해석 모델을 개발하여 해석 결과와 시험 결과를 비교 분석하였다.