• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2937-2952
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• 2021

A typical three-dimensional finite element model for a fuel assembly is established, which is composed of 16 monolithic U-10Mo fuel plates and Al alloy frame. The distribution and evolution results of temperature, displacement and stresses/strains in all the parts are numerically obtained and analyzed with a self-developed code of FUELTM. The simulation results indicate that (1) the out-of-plane displacements of Al alloy side plates are mainly attributed to the bending deformations; (2) enhanced out-of-plane displacements appear in fuel plates adjacent to the outside Al plates, which results from the occurred bending deformations due to the applied constraints of outside Al plates; (3) an intense interaction of fuel foil with the cladding occurs near the foil edge, which appears more heavily in the fuel plates adjacent to the outside Al plates. The maximum first principal stresses in the fuel foil are similar for all the fuel plates and appear near the fuel foil edge; while, the through-thickness creep strains of fuel foil in the fuel plate near the central region of fuel assembly are larger, and the induced creep damage might weaken the fuel skeleton strength and raise the fuel failure risk.

• The Journal of Engineering Geology
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• v.10 no.2
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• pp.18-35
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• 2000

The most dominant type of landslide in Korea is debris flows which mostly take place along mountain slopes during the rainy season, July to August. The landslides have been reported to begin activation when rainfall is more than 200mm within 2days. The debris flows are usually followed by translational slips which occur upper part of mountain slopes and they transit to debris flow as getting down to the valleys. Lithology, location, slope inclination, grain size distribution of soil, permeability, dry density and porosity have been proved as triggering factor causing translational slides. The triggering data taken from mapping are statistically analysed to get landslide potential quantitatively. Rock mass creeps mostly occur on well bedded sedimentary rocks in Kyeongsang Basin. Although the displacement of rock mass creep is relatively small about 1m, the creep can cause severe hazards due to relatively large volume of the involved rock mass. Examples are rock mass creep occurred in the mouth of Hwangryongsan Tunnel, in Chilgok and in Sachon in 1999. Although the direct factor of the creeps are due to slope cutting at the foot area, more attention is required A rotational slide occurring within thick soil formation or weathered rock is also closely related to bottom part of slope cutting. It is propagated circular or semi-circular type. Especially in korea, the rotational slide may be frequently occurred in Tertiary tuff area. Because they are mainly composed of volcanic ash and pyroclastic materials, well developed joints and high degree of swelling and absorption can easily cause the slide. The landslide among the Pohang-Guryongpo national road is belong to this type of slide.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.175-180
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• 1998

The U-Zr metallic alloy with low swelling HT9 cladding is the candidate for the KALIMER fuel rod. The fuel rod should be able to maintain the structural integrity during its lifetime in the reactor. In a typical metallic fuel rod, load is mainly applied by internal gas pressure, and the deformation is primarily caused by creep of the cladding. The three-dimensional FEM modelling of a fuel rod is important to predict the structural behavior in concept design stage. Using the ANSYS code, the 3-D structure analyses were performed for various configuration, element and loads. It has been shown that the present analysis model properly evaluate the structural integrity of fuel rod. The present analysis results show that the fuel rod is expected to maintain its structural integrity during normal operation.

• Journal of Auto-vehicle Safety Association
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• v.7 no.2
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• pp.19-24
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• 2015

Ferritic Nitro Carburizing (FNC) cast iron brake discs is known to improve corrosion resistance and brake creep groan noise as well as prevent corrosion-induced pulsation. But, it is necessary to treat honing machining on braking surface to avoid grinding noise during braking.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2094-2106
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• 2022

The solid-state core of a heat pipe cooled reactor operates at high temperatures over 1000 K with thermal and irradiation-induced expansion during burnup. The expansion changes the gap thickness between the solid components and the material properties, and may even cause the gap closure, which then significantly influences the thermal and mechanical characteristics of the reactor core. This study developed an irradiation behavior model for HPRTRAN, a heat pipe reactor system analysis code, to introduce the irradiation effects such as swelling and creep. The megawatt heat pipe reactor MegaPower was chosen as an application case. The coupled irradiation-thermal-mechanical model was developed to simulate the irradiation effects on the heat transfer and stresses of the whole reactor core. The results show that the irradiation deformation effect is significant, with the irradiation-induced strains up to 2.82% for fuel and 0.30% for monolith at the end of the reactor lifetime. The peak temperatures during the lifetime are 1027:3 K for the fuel and 956:2 K for monolith. The gap closure enhances the heat transfer but caused high stresses exceeding the yield strength in the monolith.

• Nuclear Engineering and Technology
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• v.9 no.4
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• pp.223-236
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• 1977

An integral computer code has been developed for a mechanical and thermal design and performance analysis of an oxide fuel rod in a pressurized water reactor. The code designated as FROD 1.0 takes into account the phenomena of radial power depression within the pellet, cracking, densification and swelling of the pellet, fission gas release, clad creep, pellet-clad contact, heat transfer to coolant and buildup of corrosion layers on the clad surface. The FROD 1.0 code yields two-dimensional temperature distributions, dimensional changes, stresses, and internal pressure of a fuel rod as a function of irradiation time within a reasonable computation time. The code may also be used for the analyses of oxide fuel rods in other thermal reactors. As an application of FROD 1.0 the behavior of fuel rod loaded in the first core of Go-ri Nuclear Power Plant Unit 1 is predicted for the two power histories corresponding to steady state operation and Codition II of the ANS Classification. The results are compared with the design criteria described in the Final Safety Analysis Report and a discrepancy between these two values is discussed herein.

• Applied Chemistry for Engineering
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• v.20 no.4
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• pp.459-463
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• 2009

In this study, physical properties of poly(vinyl acetate-co-ethylene) (VAE) emulsion were investigated by adding different amounts of di-butyl phthalate (DBP) which is a common plasticizer of VAE. The glass transition temperature $(T_g)$ of the dried plasticized VAE emulsion film, which measured by Differential Scanning Calorimeter, was decreased with increasing the DBP contents while the viscosity of the plasticized VAE emulsion was increased with the DBP contents. These results suggest that the plasticizer in the dried VAE film can prevent the strong interaction between chains, resulted by the decrease of $T_g$. In the emulsion, however, the particle sizes were swelled by the penetration of plasticizers and then its viscosity increased with the DBP content. When the DBP was added, the mechanical properties of the plasticized VAE films, such as tensile strength, elongation and creep resistance, were decreased while the water resistance was increased.

• Nuclear Engineering and Technology
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• v.20 no.2
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• pp.88-104
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• 1988

The FUel Rod Analysis(FURA) code is developed using two-dimensional finite element methods for axisymmetric and plane stress analysis of fuel rod. It predicts the thermal and mechanical behavior of fuel rod during normal and load follow operations. To evaluate the exact temperature distribution and the inner gas pressure, the radial deformation of pellet and clad, the fission gas release are considered over the full-length of fuel rod. The thermal element equation is derived using Galerkin's techniques. The displacement element equation is derived using the principle of virtual works. The mechanical analysis can accommodate various components of strain: elastic, plastic, creep and thermal strain as well as strain due to swelling, relocation and densification. The 4-node quadratic isoparametric elements are adopted, and the geometric model is confined to a half-pellet-height region with the assumption that pellet-pellet interaction is symmetrical. The pellet cracking and crack healing, pellet-cladding interaction are modelled. The Newton-Raphson iteration with an implicit algorithm is applied to perform the analysis of non-linear material behavior accurately and stably. The pellet and cladding model has been compared with both analytical solutions and experimental results. The observed and predicted results are in good agreement. The general behavior of fuel rod is calculated by axisymmetric system and the cladding behavior against radial crack is used by plane stress system. The sensitivity of strain aging of PWR fuel cladding tube due to load following is evaluated in terms of linear power, load cycle frequency and amplitude.