• Nuclear Engineering and Technology
• /
• v.22 no.1
• /
• pp.75-81
• /
• 1990

Experiments to predict the thermal contact conductance between the fuel pellet and cladding have been performed, which is important to determine the temperature distibution within the fuel rod. UO$_2$and Zircaloy-2 are used in these experiments. The measuring apparatus is composed of a presser which controls the contact pressure, a thermometer with 5.5 sheathed thermocouples, a vacuum pump, pellet and cladding rods, and two heating devices, etc. The thermal contact conductances were measured with varying the contact pressure and surface roughnesses of UO$_2$and Zircaloy-2 bars. The results show that an increase in the contact pressure and a decrease of surface roughness resulted in increase of the thermal contact conductance. Finally, a fitting correlation has been established and compared with widely-used correlations.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2002.10a
• /
• pp.305.2-305
• /
• 2002

• Journal of the Korean Ceramic Society
• /
• v.33 no.6
• /
• pp.631-636
• /
• 1996

The powder characteristics of UO2 via AUC prepared by precipitation from a UN with AC soiution produced from nuclear fuel powder conversion plant and that of the existing facility were compared. Mean particle size of AUC powder was decreased and agglomerates were much occured in case of using the AC solution that that of the gases but other properties such as particle size distribution and shape of particle are thought to be similarly. In compaction of UO2 powder the breaking pressur of agglomerated UO2 powder and the sintered density of final UO2 pellet from AC solution were measured 1.45$\times$108 N/m2 and 10.52 g/cc, These values could be used in nuclear fuel powder fabrication process.

• 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.

• Nuclear Engineering and Technology
• /
• v.46 no.1
• /
• pp.117-124
• /
• 2014

The effect of various process variables on the powder properties of recycled $U_3O_8$ from MnO-$Al_2O_3$ doped large grain $UO_2$ pellets and the effect of those recycled $U_3O_8$ powders on the sintered density and grain size of MnO-$Al_2O_3$ doped large grain $UO_2$ pellets have been investigated. The evolution of morphology, size, and BET surface area of the recycled $U_3O_8$ powders according to the respective variation of the thermo-mechanical treatment variables of oxidation temperature, powder milling, and sequential cyclic heat treatment of oxidation and then reduction was examined. The correlation between the BET surface area of recycled $U_3O_8$ powder and the sintered pellet properties of MnO-$Al_2O_3$ doped pellets showed that the pellet density and grain size of doped pellets were increased and then saturated by increasing the BET surface area of the recycled $U_3O_8$ powder. The density and grain size of the pellets were maximized when the BET surface area of the recycled $U_3O_8$ powder was in the vicinity of $3m^2/g$. Among the process variables applied in this study, the cyclic heat treatment followed by low temperature oxidation was a potential process combination to obtain the sinter-active $U_3O_8$ powder.

• Nuclear Engineering and Technology
• /
• v.34 no.5
• /
• pp.433-435
• /
• 2002

The relation between sintered densities and porosities in UO$_2$ pellets is investigated. The open porosity decreases linearly up to about 95% T.D.,(theoretical density) as the sintered density increases whereas, above 96% T.D., sintered UO$_2$ pellets do not have any open pores. The fraction of open porosity to the total porosity also decreases linearly as the sintered density increases, though the slope is lower than that of open porosity and, above 95% T.D., the fraction decreases rapidly to approach a zero.

• 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.

• Nuclear Engineering and Technology
• /
• v.55 no.6
• /
• pp.2206-2214
• /
• 2023

The existing wet reconversion processes for the recovery of scraps generated in manufacturing of nuclear fuel are complex and require several unit operation steps. In this study, it is attempted to simplify the recovery process of high-quality fuel-grade UO₂ powder. A novel wet reconversion process for uranyl nitrate hexahydrate solution is suggested by using a newly developed pulsed fluidized bed reactor, and the resultant chemical characteristics are evaluated for the intermediate ammonium uranate hydrate product and subsequently converted UO₂ powder, as well as the compliance with nuclear fuel specifications and advantages over existing wet processes. The UO₂ powder obtained by the suggested process improved fuel pellet properties compared to those derived from the existing wet conversion processes. Powder performance tests revealed that the produced UO₂ powder satisfies all specifications required for fuel pellets, including the sintered density, increase in re-sintered density, and grain size. Therefore, the processes described herein can aid realizing a simplified manufacturing process for nuclear-grade UO₂ powders that can be used for nuclear power generation.

• Proceedings of the Korean Society of Precision Engineering Conference
• /
• 1997.10a
• /
• pp.630-633
• /
• 1997

Spent fuel decladding device and dry voloxidizer is to separate the spent pellet from spent fuel rod cut by 250mm and to convert the spent pellet into powder form for reuse and/or disposal of the spent fuel. There are two methods in decladding and voloxidation of spent fuel, that is, wet method with chemical material and dry method with mechanical device. In this study, to examine the fuel rod decladding process and the pellet voloxidation process, the devices for the spent fuel decladding and the pellet voloxidation with dry method are developed. The decladding machine is designed to separate pellets from fuel rod by slitting device. And, the voloxidizer is designed to convert the spent pellet which is ceramic form into powder form by oxidation using the multi step mesh, vibrator, and air in the high temperature environment. The result of this study, such as operation condition et., will be utilized in the design of the machine for demonstration.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1977.06a
• /
• pp.112-112
• /
• 1977