Myung-Seung Yang

Robotic contamination cleaning system

This paper describes the development of a Robotic Contamination Cleaning System (RCCS) for use in the radioactive zone of the isolation room of the Irradiated Material Examination Facility (IMEF) at the Korea Atomic Energy Research Institute (KAERI). RCCS was designed to completely eliminate human interaction with the hazardous radioactive contaminants. RCCS is capable of cleaning the contaminated floor of the isolation room and collecting loose dry spent nuclear fuel debris and other radioactive waste fixed or scattered on the floor surface. RCCS is operated either by manual control or by autonomous control in conjunction with a graphical simulator, by which the human operator can monitor and intervene RCCS performing cleanup tasks in the isolation room. The mechanical design considerations, control system, and capabilities of RCCS in terms of remote cleanup operation and remote maintenance in the radioactive environment are presented.

Oxidation behavior of unirradiated UO2 pellets

Abstract UO 2 pellets were oxidised and pulverized to powder due to the volume change resulting from phase transformation of UO 2 to the intermediate phases and finally U 3 O 8 . The spallation mechanism of UO 2 pellets at 400°C was examined in this study. Transformation of UO 2 to intermediate phases was concluded to be the main driving force for the pulverization. During the formation of intermediate phases at 400°C, intergranular cracks were extensively developed, and these cracks caused the spallation. The formation of U 3 O 8 only accelerated the pulverization. The characteristics of oxidised powder were also examined.

A study of the microstructure of yttria-stabilized zirconia deposited by inductively coupled plasma spraying

Zirconia stabilized with 20 wt.% yttria was deposited to a thick free-standing type, ∼7 mm, by (inductively coupled plasma spraying) (ICPS). The spheroidization of particles and the microstructure of deposits were analyzed. Spheroidization fraction dependence on spray parameters such as particle size, H2 gas mixing quantity, and probe position was studied. Effects of parameters on the spheroidization of particles were analyzed by ANOVA (analysis of variance) (ANOVA). ANOVA results showed that the spheroidization fraction largely depend on H2 gas mixing quantity and particle size, and there are also some dependence on probe position and H2 gas mixing quantity. After melting, particles kept their chemical composition homogeneously from the center to their surface without segregation or evaporation. The degree of deformation of the diameter of the splat over the diameter of the spheroidized particle was approximately 320%, and splat thickness in the deposit varies between 2 µm and 3 µm depending on the deposition condition. The yttrium concentration gradient of the interlayer boundary appeared linear in the range of 0.5 to 1 µm. X-ray diffraction analysis and a transmission electron microscope (TEM) micrograph showed that low yttrium content particles resulted in tetragonal phase in deposit. The major characteristics of the microstructure of the thick free-standing deposit and solidification mode were studied. Microstructure of the bottom part of the deposit represented equiaxed or cellular structure. Equiaxed small grains prevailed when the droplets were quenched rapidly on substrate. The middle part of the deposit showed large columnar grains, of about 100 µm thick and 300 µm long. This may be due to high substrate or deposit temperatures and results in recrystallization and grain growth.The effects of the parameters, such as H2 gas mixing quantity, particle size, spraying distance, and probe position, on the microstructure of the deposits were evaluated. The H2 gas mixing quantity of Ar/H2=120/20 L/min compared to Ar/H2=120/10 L/min resulted in larger grain size and thicker cellular in chill. Grain shapes were affected by the heat removal rate from the deposit to its surrounding. Deposition with larger particle size showed heterogeneous grain size, insufficient particle melting, and incomplete recrystallization. The effect of probe position was less than the others.

Uranium dioxide reaction in CF4/O2 RF plasma

Abstract Research on the fluorination reaction of UO 2 in CF 4 /O 2 RF plasma is carried out at temperatures of up to 370°C under total pressure of 0.3 Torr. The reaction rates are investigated as functions of CF 4 /O 2 ratio, plasma power, substrate temperature, and exposure time to the plasma. It is found that there exists an optimum CF 4 /O 2 ratio of around four for the efficient etching, regardless of RF power and substrate temperature. According to the mass spectrometry it is revealed that the major reaction product is uranium hexa-fluoride UF 6 . Some minor species such as UF 4 and UF 5 are probably generated parasitically. The highest etching reaction rate at 370°C under 150 W exceeds 1000 monolayers/min, which is equivalent to 0.4 μm/min. Based on the experimental findings, dominant overall reaction of uranium dioxide in CF 4 /O 2 plasma is determined: UO 2 +3/2 CF 4 +3/8 O 2 = UF 6 +3/2 CO 2−x , where CO 2− x represents the undetermined mix of CO 2 and CO. The overall reaction follows a linear kinetics and is thus rate-determined by the surface reaction between the uranium atom in UO 2 F 2 intermediates on the surface and incoming fluorine atoms or fluorine containing radicals. The activation energy of this reaction is derived as 12.1 kJ/mol.

Investigation of PWR Hull with a View to Downgrade

Abstract The cladding materials remaining after the reprocessing of nuclear fuel, generally called hulls, are classified as high‐level radioactive waste. They are usually packaged in a container for disposal after being compacted, melted, or solidified into a heterogeneous matrix. Efforts to fabricate a better waste form from an environmental perspective have failed due to the technical difficulties encountered in the chemical decontamination of cladding hulls. In the early 1990s, the accumulation of radiochemical data on hulls and the advent of new technology such as laser or plasma have made the decontamination of hulls a viable option. This paper summarizes information regarding the radiochemical analysis of spent nuclear fuel hulls through a literature survey, including the characteristics of the hulls of 32,000 MWd/tU burn‐up and 15 years cooling of Korean pressurized water reactor. The reduction of the radioactivity by peeling off the inner surface of the hulls via laser technology was evaluated.

The stoichiometry and the oxygen potential change of urania fuels during irradiation

Abstract A defect model for irradiated UO 2 fuel solid-solution was devised based on the defect structure of pure urania. Using the equilibrium between fuel solid-solution and fission-products and the material balance within the fuel, the stoichiometry change of urania fuel was traced with burn-up. This tracing method was applied to high burn-up fuels. The oxygen potential of urania fuel turned out to increase slightly with burn-up. The stoichiometric change was calculated to be negligible due to the buffering role of Mo.

Characterization of irradiated simulated DUPIC fuel

A simulated DUPIC (Direct Use of Spent PWR Fuel in CANDU Reactors) fuel was irradiated at HANARO research reactor of KAERI in 1999. Post-irradiation examinations, such as measurements of γ-scanning, profilometry, density, hardness, microstructure, and fission product distribution were performed on the irradiated simulated DUPIC fuel. In γ-scanning, the intensity along the axial direction was sharply decreased at the areas between the pellets. There was no significant change in the profilometry of SEU-1.47%, but variation was detected in SEU-2.19%+F.P by 67 μm, and the peaks precisely coincided with the ridges of the pellets. The marked difference between SEU-1.47% and SEU-2.19%+F.P pellets after irradiation was the configuration of cracks arised in the pellets. Some large equiaxed grains of 11.1 μm were observed at the center of the SEU-2.19%+F.P pellet, while the grain size near the surface of the pellet was remained almost the same as the original grain size of 5.58 μm. The hardness had no tendency toward change to the direction, but average hardness was increased as much as 10% compared with a fresh simulated DUPIC fuel.

Equipment Arrangements for DUPIC Nuclear Fuel Fabrication and Their Remote Operation and Maintenance in Hot-Cell

DUPIC (Direct Use of spent PWR fuel In CANDU reactors) nuclear fuel that reuses spent PWR (Pressurized Water Reactor) fuel as raw material is being developed at KAERI (Korea Atomic Energy Research Institute). All DUPIC nuclear fuel fabrication processes that mainly consist of powder preparation, pelletization and rod fabrication are conducted in the completely shielded M6 hot-cell of the IMEF (Irradiated Material Examination Facility) at KAERI because of the nature of the high radioactivity of spent PWR fuel. Various types of equipment specially designed for DUPIC fuel fabrication are operated and maintained in a remote manner. This paper presents the in-cell arrangements of all the equipment used for DUPIC nuclear fuel fabrication in the confined hot-cell facility and their remote operation and maintenance in situ.

Mitigation of Radiation Load by Trapping of Tritium in Off-gas during Dry-processes

Although tritium gas is produced a small amount comparing to the other fission products, it should be controlled for long half-life, high residence time, high isotopic exchange rate and ease of assimilation into living matter. A demand for lengthening the life time of disposal site needs treatment of spent nuclear fuel to reduce the volume of high level radwaste. This study is for trapping the tritium gas during dry processes carried out in oxygen condition, having a potential of exposure into operators and environment. The experiments were performed into two different hydrogen concentration ranges, i.e. 1000 ~ 9000 ppm of high range of hydrogen concentration and 25 ~ 100 ppm of low range of hydrogen concentration. At high range of hydrogen concentration, H2O conversion ratio at 400°C indicated above 98 % up to 7000 ppm, and 100% at 450°C. All the results of H2O conversion ratio at the low range of hydrogen concentration are represented close to 99%. Converted H2O vapor was adsorbed at the Molecular sieve close to 100%. More than 98% of H2O conversion ratio was attained up to 4 cm/s of linear gas velocity, whereas over 99% for the low range of hydrogen concentration. Some catalytic effect of ability of conversion hydrogen into H2O on stainless steel was studied.

Hot-Cell Shielding System for High Power Transmission in DUPIC Fuel Fabrication

This paper presents a newly designed hot-cell shielding system for use in the development of DUPIC (Direct Use of spent PWR fuel In CANDU reactors) fuel at KAERI (Korea Atomic Energy Research Institute). This hot-cell shielding system that was designed to transmit high power to sintering furnace in-cell from the out-of-cell through a thick cell wall has three subsystems - a steel shield plug with embedded spiral cooling line, stepped copper bus bars, and a shielding lead block. The dose-equivalent rates of the hot-cell shielding system and of the apertures between this system and the hot-cell wall were calculated. Calculated results were compared with the allowable dose limit of the existing hot-cell. Experiments for examining the temperature changes of the shielding system developed during normal furnace operation were also carried out. Finally, gamma-ray radiation survey experiments were conducted by Co-60 source. It is demonstrated that, from both calculated and experimental results, the newly designed hot-cell shielding system meets all the shielding requirements of the existing hot-cell facility, enabling high power transmission to the in-cell sintering furnace.