• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(2)
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• pp.171-176
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• 1997

UO$_2$-Gd$_2$O$_3$fuel has been sintered to study the effect of powder property and sintering atmospheres on densification and microstructure. Three types of powders have been used; AUC-UO$_2$ powder and ADU-UO$_2$ powder were mixed with Gd$_2$O$_3$ Powder, and co-milled AUC-UO$_2$ and Gd$_2$O$_3$ powder. UO$_2$-(2, 5, 10)wt% Gd$_2$O$_3$pellets have been sintered at 168$0^{\circ}C$ for 4 hours in the mixture of H$_2$ and $CO_2$ gases, of which oxygen potential has been controlled by the ratio of $CO_2$ to H$_2$ gas. Densities of UO$_2$-Gd$_2$O$_3$ fuel pellets are quite dependent on powder types, and UO$_2$-Gd$_2$O$_3$ fuel using co-milled UO$_2$ powder yields the highest density. A long range homogeneity of Gd is determined by powder mixing. As the oxygen potential of sintered atmosphere increases, the sintered densities of UO$_2$-Gd$_2$O$_3$ pellets decrease but grain size increases. In addition, (U, Gd)O$_2$ solid solution becomes more homogeneous. The UO$_2$-Gd$_2$O$_3$fuel having adequate density and homogeneous microstructure can be fabricated by co-milling powder and by high oxygen potential.

• 한국분말재료학회지
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• 제16권4호
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• pp.254-261
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• 2009

In this study, we investigated the unit process parameters in spherical $UO_2$ kernel preparation. Nearly perfect spherical $UO_3$ microspheres were obtained from the 0.6M of U-concentration in the broth solution, and the microstructure of the $UO_2$ kernel appeared the good results in the calcining, reducing, and sintering processes. For good sphericity, high density, suitable microstructure, and no-crack final $UO_2$ microspheres, the temperature control range in calcination process was $300{\sim}450^{\circ}C$, and the microstructure, the pore structure, and the density of $UO_2$ kernel could be controlled in this temperature range. Also, the concentration changes of the ageing solution in aging step were not effective factor in the gelation of the liquid droplets, but the temperature change of the ageing solution was very sensitive for the final ADU gel particles.

• Nuclear Engineering and Technology
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• 제43권2호
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• pp.141-148
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• 2011

This paper presents the quantitative analysis results of research on the burnup effect on the nuclear fuel cycle cost of BeO-$UO_2$ fuel. As a result of this analysis, if the burnup is 60 MWD/kg, which is the limit under South Korean regulations, the nuclear fuel cycle cost is 4.47 mills/kWh at 4.8wt% of Be content for the BeO-$UO_2$ fuel. It is, however, reduced to 3.70 mills/kWh at 5.4wt% of Be content if the burnup is 75MWD/kg. Therefore, it seems very advantageous, in terms of the economic aspect, to develop BeO-$UO_2$ fuel, which does not have any technical problem with its safety and is a high burnup & long life cycle nuclear fuel.

• 한국세라믹학회지
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• 제46권6호
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• pp.574-582
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• 2009

$UO_2$ microsphere particles, core material of HTGR(High Temperature Gas Reactor) nuclear fuel, were prepared using by the GSP(Gel Supported Precipitation) method which is a modified-method of the wet sol-gel process. The spherical shape of initial liquid ADU droplets from the vibration nozzle system was continuously kept till the conversion to the final $UO_2$ microsphere. But the size of a final $UO_2$ microsphere was shrunken to about 25% of an initial ADU droplet size. Also, we found that the composition of dried-ADU gel particles was constituted of the very complicated phases, coexisted the U=O, C-H, N-H, N-O, and O-H functional groups by FT-IR. The important factors for obtain the no-crack $UO_2$ microsphere during the thermal treatment processes must perfectly wash out the remained-$NH_4NO_3$ within the ADU gel particle in washing process and the selections of an appropriate heating rate at a suitable gas atmosphere, during the calcining of ADU gel particles, the reducing of $UO_3$ particles, and the sintering of $UO_2$ particles, respectively.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.1002-1008
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• 2016

An understanding of the coarsening process of the large fission gas pores in the high burn-up structure (HBS) of irradiated $UO_2$ fuel is very necessary for analyzing the safety and reliability of fuel rods in a reactor. A numerical model for the description of pore coarsening in the HBS based on the Ostwald ripening mechanism, which has successfully explained the coarsening process of precipitates in solids is developed. In this model, the fission gas atoms are treated as the special precipitates in the irradiated $UO_2$ fuel matrix. The calculated results indicate that the significant pore coarsening and mean pore density decrease in the HBS occur upon surpassing a local burn-up of 100 GWd/tM. The capability of this model is successfully validated against irradiation experiments of $UO_2$ fuel, in which the average pore radius, pore density, and porosity are directly measured as functions of local burn-up. Comparisons with experimental data show that, when the local burn-up exceeds 100 GWd/tM, the calculated results agree well with the measured data.

• 한국세라믹학회지
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• 제47권6호
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• pp.590-597
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• 2010

HTGR using a TRISO coated particles as nuclear raw fuel material can be used to produce clean hydrogen gas and process heat for a next-generation energy source. For these purposes, a TRISO coated particle was prepared with 3 pyro-carbon (buffer, IPyC, and OPyC) layers and 1 silicone carbide (SiC) layer using a CVD technique on a spherical $UO_2$ kernel surface as a fissile material. In this study, a spherical $UO_2$ particle was prepared using a modified sol-gel method with a vibrating nozzle system, and TRISO coating fabrication was carried out using a fluidized bed reactor with coating gases, such as acetylene, propylene, and methyltrichlorosilane (MTS). As the results of this study, a spherical $UO_2$ kernel with a sphericity of 1+0.06 was obtained, and the main process parameters in the $UO_2$ kernel preparation were the well-formed nature of the spherical ADU liquid droplets and the suitable temperature control in the thermal treatment of intermediate compounds in the ADU, $UO_3$, and $UO_2$ conversions. Also, the important parameters for the TRISO coating procedure were the coating temperature and feed rate of the feeding gas in the PyC layer coating, the coating temperature, and the volume fraction of the reactant and inert gases in the SiC deposition.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.11-18
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• 2022

In this research, curium oxide (CmO₂) is studied as fuel for VVER-1200 reactor to get an attention to its energy value and possibilities. For this purpose, CmO₂ is used in fuel rods or integrated burnable absorber (IBA) rods with and without UO₂ and then compared with the conventional fuel assembly of VVER-1200 reactor. It is burned to 60 GWd/t by using SRAC-2006 code and JENDL-4.0 data library. From these studies, it is found that CmO₂ is competent like UO₂ as a fuel due to higher fission cross-section of ²⁴³Cm and ²⁴⁵Cm isotopes and neutron capture cross-section of ²⁴⁴Cm and ²⁴⁶Cm isotopes. As a result, when some or all of the UO₂ of fuel rods or IBA rods are replaced by CmO₂, we get a similar k-inf like the reference even with lower enrichment UO₂ fuels. These studies show that the use of CmO₂ as IBA rods is more effective than the fuel rods considering the initially loaded amount, power peaking factor (PPF), fuel temperature and void coefficient, and the quality of spent fuel. From a detailed study, 3% CmO₂ with inert material ZrO₂ in IBA rods are recommended for the VVER-1200 reactor assembly from the once through concept.

• 한국세라믹학회지
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• 제33권6호
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• pp.631-636
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• 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
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• 제56권1호
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• pp.167-179
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• 2024

This study explores the feasibility of employing (U, Th)-based accident tolerant fuels (ATFs), specifically (_0.8UO₂, _0.2ThO₂), (_0.8UN, _0.2ThN), and (_0.8UC, _0.2ThC). The investigation assesses the overall performance of these proposed fuel materials in comparison to the conventional UO₂, focusing on deep neutronic and thermal-hydraulic (Th) analyses. Neutronic analysis utilized the MCNPX code, while COMSOL Multiphysics was employed for thermal-hydraulic analysis. The primary objective of this research is to overcome the limitations associated with traditional UO₂ fuel by exploring alternative fuel materials that offer advantages in terms of abundance and potential improvements in performance and safety. Given the limited abundance of UO₂, long-term sustainable nuclear energy production faces challenges. From a neutronic standpoint, the U-Th based fuels demonstrated remarkable fuel cycle lengths, except (_0.8UN, _0.2ThN), which exhibited the minimum fuel cycle length and, consequently, the lowest fuel burn-up. Regarding thermal-hydraulic performance, (_0.8UN, _0.2ThN) exhibited outstanding performance with significant margins against fuel melting compared to the other materials. Overall, when considering the integrated performance, the most favourable results were obtained with the use of the (_0.8UC, _0.2ThC) fuel configurations. This study contributes valuable insights into the potential benefits of (U, Th)-based ATFs as a promising avenue for enhanced nuclear fuel performance.

• Nuclear Engineering and Technology
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• 제30권2호
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• pp.128-139
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• 1998

The effect of UO$_2$ powder property and oxygen potential on characteristics of sintered UO$_2$-Gd$_2$O$_3$ fuel pellets has been investigated. Two types of powder, mixture of AUC-UO$_2$ and Gd$_2$O$_3$powders (type I) and mixture of ADU-UO$_2$ and Gd$_2$O$_3$powders (type II), have been prepared, pressed, and sintered at 168$0^{\circ}C$ for 4 hours. Four sintering atmospheres with different mixing ratios of $CO_2$to H$_2$ gas ranging from 0 to 0.3 have been used. UO$_2$-Gd$_2$O$_3$ fuel has lower sintered density than UO$_2$ fuel, and the density drop is larger for powder type I than for powder type II. As the oxygen potential increases, the sintered density of UO$_2$-2wt% Gd$_2$O$_3$pellets increases but that of UO$_2$-10wt% Gd$_2$O$_3$ pellets decreases. It is found that pores are newly formed in UO$_2$-10wt% Gd$_2$O$_3$ pellets in accordance with the decrease in density. The grain size of UO$_2$-Gd$_2$O$_3$ fuel increases and a short range G4 distribution becomes homogeneous as the oxygen potential increases. A long range ed distribution and grain structure are inhomogeneous for powder type II. The lattice parameter of (U,Gd)O$_2$solid solution decreases linearly with Gd$_2$O$_3$ content. The dependence of UO$_2$-Gd$_2$O$_3$fuel characteristics on powder type and sintering atmosphere have been discussed.