• Nuclear Engineering and Technology
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• 제45권4호
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• pp.565-572
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• 2013

The fabrication methods and requirements of the fiber, interphase, and matrix of nuclear grade $SiC_f/SiC$ composites are briefly reviewed. A CVI-processed $SiC_f/SiC$ composite with a PyC or $(PyC-SiC)_n$ interphase utilizing Hi-Nicalon Type S or Tyranno SA3 fiber is currently the best combination in terms of the irradiation performance. We also describe important material issues for the application of SiC composites to LWR fuel cladding. The kinetics of the SiC corrosion under LWR conditions needs to be clarified to confirm the possibility of a burn-up extension and the cost-benefit effect of the SiC composite cladding. In addition, the development of end-plug joining technology and fission products retention capability of the ceramic composite tube would be key challenges for the successful application of SiC composite cladding.

• Tribology and Lubricants
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• 제30권6호
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• pp.343-349
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• 2014

Currently, in order to resist the scratching of the fuel rod surface while fabricating the fuel assembly of the light-water nuclear reactor, we use a solution of nitrocellulose, an explosive material, as a dry lubricant along with its solvent. However, the demand for developing safe and harmless aqueous alternative materials for environment-conservation and field-worker safety has increased. In this study, we demonstrate the preparation of a novel aqueous resin composite using a formulation of aqueous polymeric resin, alcoholic solvent, and water. Subsequently, we characterize this composite on the basis of hardness, adhesive property, and water solubility using plates similar to the fuel rod material. The insertion test of a fuel rod coated with the YS-3 composite shows load values of $18.8-20.5kg/cm^2$, which is comparable with $18.8-20.5kg/cm^2$ of the nitrocellulose coating agent. In addition, the depth and width of longitudinal scratches caused by the YS-3 composite test are 50% higher than those of the standard. We can develop a harmless and safe aqueous dry lubricant to replace the existing NC products through field testing of 264 pieces of fuel rods, after producing 350 kg of the YS-3 prototype. The scratch test for the rod surface showed that weight of chip of YS-3 prototype was smaller than that of NC before and after solvent treatment, indicating the properties of YS-3 prototype was comparable to the counterpart.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3725-3731
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• 2023

Thorium dioxide (ThO₂)-silicon carbide (SiC) composite fuel pellets were fabricated via the spark plasma-sintering (SPS) method to investigate the role of the addition of SiC in enhancing the thermal conductivity of ThO₂ fuel. SiC particles with an average size of 1㎛ in 10 and 15 vol% were used to manufacture the composite pellets. The changes in the composites' densification, microstructure and thermal conductivity were explored by comparing them with pure ThO₂ pellets. The structural and microstructural characterization of the composite pellets has revealed that SPS could manufacture high-quality composite pellets without having any reaction products or intermetallic. The density measurement by the Archimedes principles and the grain size from the electron back-scattered diffraction (EBSD) analysis has indicated that the composites have higher densities and smaller grain sizes than the pellets without SiC addition. Finally, thermal conductivity as a function of temperature has revealed that sintered ThO₂-SiC composites showed an increase of up to 56% in thermal conductivity compared to pristine ThO₂ pellets.

• 한국세라믹학회지
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• 제51권5호
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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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• 제53권8호
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• pp.2682-2695
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• 2021

This paper aims to evaluate the structural dynamic responses and damage/failure of the nuclear fuel reprocessing plant under the free drop impact of spent fuel cask (SFC) and fuel assembly (FA) during the on-site transportation. At the present Part I of this paper, the large-scale SFC model free drop test and the corresponding numerical simulations are performed. Firstly, a composite target which is composed of the protective structure, i.e., a thin RC plate (representing the inverted U-shaped slab in the loading shaft) and/or an autoclaved aerated concrete (AAC) blocks sacrificial layer, as well as a thick RC plate (representing the bottom slab in the loading shaft) is designed and fabricated. Then, based on the large dropping tower, the free drop test of large-scale SFC model with the mass of 3 t is carried out from the height of 7 m-11 m. It indicates that the bottom slab in the loading shaft could not resist the free drop impact of SFC. The composite protective structure can effectively reduce the damage and vibrations of the bottom slab, and the inverted U-shaped slab could relieve the damage of the AAC blocks layer dramatically. Furthermore, based on the finite element (FE) program LS-DYNA, the corresponding refined numerical simulations are performed. By comparing the experimental and numerical damage and vibration accelerations of the composite structures, the present adopted numerical algorithms, constitutive models and parameters are validated, which will be applied in the further assessment of drop impact effects of full-scale SFC and FA on prototype nuclear fuel reprocessing plant in the next Part II of this paper.

• Nuclear Engineering and Technology
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• 제48권1호
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• pp.1-15
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• 2016

For a long time, a top priority in the nuclear industry was the safe, reliable, and economic operation of light water reactors. However, the development of accident-tolerant fuel (ATF) became a hot topic in the nuclear research field after the March 2011 events at Fukushima, Japan. In Korea, innovative concepts of ATF have been developing to increase fuel safety and reliability during normal operations, operational transients, and also accident events. The microcell $UO_2$ and high-density composite pellet concepts are being developed as ATF pellets. A microcell $UO_2$ pellet is envisaged to have the enhanced retention capabilities of highly radioactive and corrosive fission products. High-density pellets are expected to be used in combination with the particular ATF cladding concepts. Two concepts-surface-modified Zr-based alloy and SiC composite material-are being developed as ATF cladding, as these innovative concepts can effectively suppress hydrogen explosions and the release of radionuclides into the environment.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1330-1338
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• 2018

High-density UMo/Zr monolithic nuclear fuel plates have a promising application prospect in high flux research and test reactors. The solid state welding method called co-rolling is used for their fabrication. Hot co-rolling simulations for the composite blanks of UMo/Zr monolithic nuclear fuel plates are performed. The effects of coupon sizes and mechanical property parameters on the contact pressures between the to-be-bonded surfaces are investigated and analyzed. The numerical simulation results indicate that 1) the maximum contact pressures between the fuel coupon and the Zircaloy cover exist near the central line along the plate length direction; as a whole the contact pressures decrease toward the edges in the plate width direction; and lower contact pressures appear at a large zone near the coupon corner, where de-bonding is easy to take place in the in-pile irradiation environments; 2) the maximum contact pressures between the fuel coupon and the Zircaloy parts increase with the initial coupon thickness; after reaching a certain thickness value, the contact pressures hardly change, which was mainly induced by the complex deformation mechanism and special mechanical constitutive relation of fuel coupon; 3) softer fuel coupon will result in lower contact pressures and form interfaces being more out-of-flatness.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.4052-4059
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• 2021

The microstructure of the fuel pellet plays an essential role in fission gas buildup and release and is critical for the safe and continued operation of nuclear power stations. Structural analysis of uranium dioxide (UO₂)-molybdenum (Mo) composite fuel pellets prepared at a range of sintering temperatures from 1300 to 1800 ℃ was performed. Mo micro and nanoparticles were used in making the composite pellets. A systematic investigation into the influence of processing parameters during Spark Plasma Sintering (SPS) of the pellets on the microstructure, texture, grain size, and grain boundary characters of UO₂-Mo is presented. UO₂-Mo composite show significant differences in the fraction of general boundaries and also special/coincident site lattice (CSL) boundaries. EBSD orientation maps demonstrated that <111> texturing was observed in the pellets fabricated at 1500 ℃. The experimental investigations suggest that UO₂-Mo composite pellets have favorable microstructural features compared to the UO₂ pellet.

• 한국세라믹학회지
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• 제51권5호
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• pp.435-441
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• 2014

Silicon carbide-based ceramics and their composites have been studied for application to fusion and advanced fission energy systems. For fission reactors, $SiC_f$/SiC composites can be applied to core structural materials. Multilayered SiC composite fuel cladding, owing to its superior high temperature strength and low hydrogen generation under severe accident conditions, is a candidate for the replacement of zirconium alloy cladding. The SiC composite cladding has to retain its mechanical properties and original structure under the inner pressure caused by fission products; as such it can be applied as a cladding in fission reactor. A hoop strength test using an expandable polyurethane plug was designed in order to evaluate the mechanical properties of the fuel cladding. In this paper, a hoop strength test of the multilayered SiC composite tube for nuclear fuel cladding was simulated using FEA. The stress caused by the plug was distributed nonuniformly because of the friction coefficient difference between the inner surface of the tube and the plug. Hoop stress and shear stress at the tube was evaluated and the relationship between the concentrated stress at the inner layer of the tube and the fracture behavior of the tube was investigated.

• Composites Research
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• 제35권3호
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• pp.161-174
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• 2022

SiC 섬유강화 복합체는 경수형 원자로의 안전성을 획기적으로 향상시킬 수 있는 사고저항성 핵연료 피복관 소재이다. 지르코늄 합금 피복관 및 금속기반 사고저항성 핵연료 피복관에 비해, 중대 사고 환경에서도 우수한 구조적 안정성을 가지고 부식 속도가 매우 낮아, 사고 시 원자로의 온도를 낮추고 사고 진행을 늦출 수 있다. 본 논문에서는 현재 개발되고 있는 사고저항성 SiC 복합체 핵연료 피복관의 개념 및 가동/사고환경에서의 다양한 특성, 상용화를 위해 해결해야 할 다양한 이슈에 대해서 소개하고자 한다.