Heemoon Kim

Xenon Diffusivity in Thoria-Urania Fuel

Abstract Postirradiation annealing tests were performed to obtain the 133Xe diffusion coefficients in uranium dioxide (UO2) and mixed thorium-uranium dioxide [(Th-U)O2] fuels. Specimens were a single-grained UO2, a polycrystalline UO2, and a polycrystalline (Th-U)O2. The (Th-U)O2 specimen was a mixture of 35% ThO2 and 65% UO2. Each 300-mg specimen was irradiated to a burnup of 0.1 MWd/t U. Postirradiation annealing tests were performed at 1400, 1500, and 1600°C, continuously. The xenon diffusion coefficients for the nearly stoichiometric single-grained UO2 agree well with the data of others. The xenon diffusion coefficients in the polycrystalline (Th-U)O2 are approximately one order lower than those in the polycrystalline UO2. The xenon diffusion coefficient in the (Th-U)O2 increases with the increasing oxygen potential of the ambient gas.

Measurement of microstructure and eutectic penetration rate on irradiated metallic fuel after high-temperature heating test

Microstructural development of irradiated U-10Zr fuel slug with T92 cladding specimen was examined after thermal exposure of 750 °C for 1 h. Optical microscopy, scanning electron microcopy and electron microprobe analysis were employed to examine the microstructure, constituent migration, and eutectic penetration. Migration phenomena of U, Zr, Fe, and Cr indicative of Soret effect was observed, and Nd lanthanide fission product was found at the eutectic melting region. Eutectic penetration was quantified and correlated to a thermal activation model with a good agreement. Compared to the previously reported eutectic penetration rates for the unirradiated U-10Zr fuel slug with FMS (Ferritic Martensitic Steel, HT9) cladding specimens, the eutectic penetration rate determined from this study for the irradiated fuel specimen was higher. This phenomenon can be caused by the effect of lanthanide fission product migration into fuel slug-cladding interface during irradiation, and lowering the eutectic threshold temperature for the irradiated fuel.

Shielding Considerations for Advanced Fuel Irradiation Experiments

An in-pile test program for the development of a high burn-up fuel is planned for the HANARO reactor. The source term originates from a leakage of fission products from the anticipated failed fuels into the gas flow tubes and around the instrumentation and control system. In order to quantify the fuel composition in the event of a fuel failure, the isotope generation and depletion code ORIGEN 2.0 was used. The computer program Microshield 6.2 was used to calculate the doses from specific locations, where a high radioactivity is expected during an irradiation. The results indicate that the equivalent dose in the investigated working areas is less than the permitted dose rate of 6.25 μSv/hr. However, access to the area of a decay vessel may need to be limited, and the installation of a Pb wall with a 20.5 cm thickness is recommended. From the analysis of a radioactive decay with time, most of the concerned gaseous nuclides with short half-lives after 3 months, were decayed, with one exception which was Kr-85, thus it should be released in accordance with applicable government laws after measuring its activity in individual holding vessels.

Diffusion Coefficients of Volatile Fission Products in Single Crystal Urania

The diffusion coefficient of Xe-133 was obtained from an annealing test. The specimen was made from U02 single crystal powder with natural enrichment and a grain size of 25 μm. Gamma scanning was performed before and after the annealing test to obtain total fission gas release fractions. The annealing test was performed for 10 hours with three temperature changes. Gamma ray peaks from the trap system were measured every 20 minutes during the annealing test. Activation energy and the pre-exponential factor of the diffusion coefficient equation were calculated using the release fraction data and a diffusion equation. The values were determined to be 360(±0.5%) kJ/mol and 2.0xl0-7(±10%), respectively.