Kun-Woo Song

KAERI’s Development of LWR Accident-Tolerant Fuel

Abstract The Fukushima accident has had a tremendous impact on Japan and the rest of the world in the areas of public health, economy, and nuclear energy policy. Thus, international consensus has been reached that inherent tolerance of nuclear fuel to severe accidents needs to be increased significantly to prevent accidents or to mitigate their consequences. In this respect, several countries have started to develop accident-tolerant fuel (ATF) that can tolerate loss of active cooling for a considerably longer time period than current fuels, while maintaining or improving performance during normal operations and operational transients and also enhancing fuel safety for beyond-design-basis events. The Korea Atomic Energy Research Institute is also developing ATF: surface-coated Zr cladding and metal-ceramic hybrid cladding for the purpose of suppressing hydrogen generation during severe accidents, and microcell UO2 pellets to enhance the retention of highly radioactive and corrosive fission products such as Cs and I, where all UO2 grains are enveloped by thin cell walls that act as chemical traps or physical barriers for the movement of fission products. When the screening of developing fuel materials has been performed through various out-of-pile tests, irradiation tests of the selected materials will be carried out in a research reactor to demonstrate their enhanced accident tolerance.

THERMOMECHANICAL ANALYSIS AND IRRADIATION TEST OF SINTERED DUAL-COOLED ANNULAR FUEL

This paper presents some of the key technologies in the area of fuel performance that Korea Atomic Energy Research Institute (KAERI) has developed for a dual-cooled annular fuel, which should be available before the annular fuel can be considered to be used in a commercial nuclear power plant. First, considering the characteristics of the annular fuel - that it has two coolant channels, outer and inner, and also two gaps between the pellet and cladding - KAERI has developed a computer code DUOS that calculates temperature, swelling, densification, and stress and strain in the annular fuel. The DUOS code was verified by comparing it with either ABAQUS or analytical solutions. The first irradiation test of sintered annular fuel pellets with different initial densities was performed in the HANARO reactor up to a pellet burnup of 10.9 MWd/kg U and then subjected to postirradiation examination. Gamma scanning along the axial direction of the irradiated fuel rods showed the geometrical integrity of the annular fuel pellets, ruling out the possibility that fragmented annular pellet cracks could move down along the axial direction of the fuel rod and hence the pellet stack length could be reduced. Macroscopy of the annular fuel pellets revealed many radial and circumferential cracks that could lead to different outer and inner gap sizes along the axial direction of the annular fuel rod, which would suggest that heat transfer to both the outer and inner coolant channels during the irradiation of annular fuel rods would depend on the axial profile of the two gaps along the axial direction. The swelling rate derived from density measurement of the annular fuel pellets with 98.0% theoretical density was 0.25 to 0.60 vol % per 10 MWd/kg U, corresponding to the one observed for solid fuel pellets irradiated at low temperature.

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.