• Journal of the Korean Ceramic Society
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• v.51 no.5
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• pp.418-423
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• 2014

Fully ceramic micro encapsulated (FCM) nuclear fuel is a concept recently proposed for enhancing the stability of nuclear fuel. FCM nuclear fuel consists of tristructural-isotropic (TRISO) fuel particles within a SiC matrix. Each TRISO fuel particle is composed of a $UO_2$ kernel and a PyC/SiC/PyC tri-layer which protects the kernel. The SiC ceramic matrix is created by sintering. In this FCM fuel concept, fission products are protected twice, by the TRISO coating layer and by the SiC ceramic. The SiC ceramic has proven attractive for fuel applications owing to its low neutron-absorption cross-section, excellent irradiation resistivity, and high thermal conductivity. In this study, a SiC-matrix composite containing TRISO particles was sintered by hot pressing with $Al_2O_3-Y_2O_3$ additive system. Various sintering conditions were investigated to obtain a relative density greater than 95%. The internal distribution of TRISO particles within the SiC-matrix composite was observed using an x-ray radiograph. The fracture of the TRISO particles was investigated by means of analysis of the cross-section of the SiC-matrix composite.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.1-12
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• 2019

One of proposed Accident Tolerant Fuel (ATF) concept is fully ceramic micro-encapsulated fuel (FCMF). FCMF using uranium mononitride (UN) has better safety aspects than $UO_2$ pellet fuel although it might not have a better neutronic performance due to the presence of matrix and high neutron-induced interaction of $^{14}N$. Before implementing UN-FCMF technology in Small-PWR, further research must be taken place to make sure the proposed design of fuel assembly has inherent safety features and maintain the fuel performance. This study focusses on the neutronic analysis of UN-FCMF based fuel assembly using Serpent and SCALE codes. It is shown in the proposed fuel assembly design has inherent safety features with respect to the fuel temperature reactivity coefficient, void reactivity coefficient, and moderator temperature reactivity coefficient. It is noted that the use of FCMF leads to a lower ratio of burnup to $^{235}U$ enrichment ratio compared to the $UO_2/Zr$ fuel.

• Journal of the Korean Ceramic Society
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• v.55 no.4
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• pp.400-405
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• 2018

FCM nuclear fuel, a concept proposed as an accident tolerant fuel in light water reactors, consists of TRISO fuel particles embedded in a SiC matrix. The uniform dispersion of internal TRISO fuel particles in the FCM fuel is very important for improving the fuel efficiency. In this study, FCM sintered pellets with various volume ratios of TRISO-coated particles were prepared by hot press sintering. The distribution of TRISO-coated particles was quantitatively analyzed using X-ray ${\mu}CT$ and expressed as a dispersion uniformity index. TRISO-coated particles were most uniformly dispersed in the FCM pellets prepared using only overcoated TRISO particles without mixing of additional SiC matrix powder. FCM pellets with uniformly dispersed TRISO particle volume fraction of up to 50% were prepared using overcoated TRISO particles with varying thickness.

• Journal of the Korean Ceramic Society
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• v.33 no.8
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• pp.953-959
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• 1996

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.650-659
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• 2016

As a type of accident-tolerant fuel, fully ceramic microencapsulated (FCM) fuel was proposed after the Fukushima accident in Japan. The FCM fuel consists of tristructural isotropic particles randomly dispersed in a silicon carbide (SiC) matrix. For a fuel element with such high heterogeneity, we have proposed a two-temperature homogenized model using the particle transport Monte Carlo method for the heat conduction problem. This model distinguishes between fuel-kernel and SiC matrix temperatures. Moreover, the obtained temperature profiles are more realistic than those of other models. In Part I of the paper, homogenized parameters for the FCM fuel in which tristructural isotropic particles are randomly dispersed in the fine lattice stochastic structure are obtained by (1) matching steady-state analytic solutions of the model with the results of particle transport Monte Carlo method for heat conduction problems, and (2) preserving total enthalpies in fuel kernels and SiC matrix. The homogenized parameters have two desirable properties: (1) they are insensitive to boundary conditions such as coolant bulk temperatures and thickness of cladding, and (2) they are independent of operating power density. By performing the Monte Carlo calculations with the temperature-dependent thermal properties of the constituent materials of the FCM fuel, temperature-dependent homogenized parameters are obtained.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.660-672
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• 2016

In Part I of this paper, the two-temperature homogenized model for the fully ceramic microencapsulated fuel, in which tristructural isotropic particles are randomly dispersed in a fine lattice stochastic structure, was discussed. In this model, the fuel-kernel and silicon carbide matrix temperatures are distinguished. Moreover, the obtained temperature profiles are more realistic than those obtained using other models. Using the temperature-dependent thermal conductivities of uranium nitride and the silicon carbide matrix, temperature-dependent homogenized parameters were obtained. In Part II of the paper, coupled with the COREDAX code, a reactor core loaded by fully ceramic microencapsulated fuel in which tristructural isotropic particles are randomly dispersed in the fine lattice stochastic structure is analyzed via a two-temperature homogenized model at steady and transient states. The results are compared with those from harmonic- and volumetric-average thermal conductivity models; i.e., we compare $k_{eff}$ eigenvalues, power distributions, and temperature profiles in the hottest single channel at a steady state. At transient states, we compare total power, average energy deposition, and maximum temperatures in the hottest single channel obtained by the different thermal analysis models. The different thermal analysis models and the availability of fuel-kernel temperatures in the two-temperature homogenized model for Doppler temperature feedback lead to significant differences.

• Nuclear Engineering and Technology
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• v.50 no.2
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• pp.253-258
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• 2018

A microcell $UO_2$ pellet, as an accident-tolerant fuel pellet, is being developed to enhance the accident tolerance of nuclear fuels under accident conditions as well as the fuel performance under normal operation conditions. Improved capture-ability for highly radioactive and corrosive fission product (Cs and I) is the distinct feature of a ceramic microcell $UO_2$ pellet, and the enhanced pellet thermal conductivity is that of a metallic microcell $UO_2$ pellet. The fuel temperature can be effectively decreased by enhanced thermal conductivity. In this study, the material concepts of metallic and ceramic microcell $UO_2$ pellets were designed, and the fabrication process of microcell $UO_2$ pellets embodying the designed concept was developed. We successfully implemented the microcell $UO_2$ pellets and produced microcell $UO_2$ pellets. In addition, an assessment of the out-of-pile properties of a microcell $UO_2$ pellet was performed, and the in-reactor performance and behavior of the developed microcell pellets were evaluated through a Halden irradiation test. According to the expectations, the excellent performance of the microcell $UO_2$ pellets was confirmed by the online measurement data of the Halden irradiation test.

• Journal of the Korean Ceramic Society
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• v.40 no.5
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• pp.410-414
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• 2003

Characteristics of $UO_{2+x}$ powders oxidized at different temperatures were examined. Pellets were fabricated by adding these oxidation powders and their properties were also investigated. Particle size of the $UO_{2+x}$ powders decreased with increasing oxidation temperature while surface area increased. Only the powders oxidized at 35$0^{\circ}C$ enhanced the strength of green pellet. However, 35$0^{\circ}C$ oxidized powders added pellet had many surface defects. The difference of shrinkage rate between the oxidized and UO$_2$ powders was thought to be the cause of them.

• Journal of the Korean Ceramic Society
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• v.45 no.11
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• pp.725-733
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• 2008

In this study the preparation method of the spherical ADU droplets, intermediate compound of a HTGR nuclear fuel, was detailed-reviewed and then, the characteristics on an ageing and a washing steps among the wet process and the thermal treatment process on the died-ADU${\rightarrow}UO_3$ conversion with the high temperature furnaces were studied. The key parameters for spherical droplets forming are a precise control of feed rate and a suitable viscosity value selection of a broth solution. Also, a harmony of vibrating frequency and amplitude of a vibration dropping system are important factor. In our case, an uranium concentration is $0.5{\sim}0.7mol/l$, viscosity is $50{\sim}80$ centi-Poise, vibration frequency is about 100Hz. In thermal treatment for no crack spherical $UO_3$ particle, the heating rate in the calcination must be operated below $2^{\circ}C$/min, in air atmosphere.

• Journal of the Korean Ceramic Society
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• v.33 no.6
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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.