• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2629-2639
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• 2021

Within the frame of the EMPIrE test, four monolithic mini-plates were irradiated in the ATR reactor. In two of them, the monolithic U(Mo) foil had been PVD-coated with Zr before the plate manufacturing. Extensive microstructural characterizations were performed on a fresh archive mini-plate, using Optical Microscopy (OM), Scanning Electron Microscopy (SEM) combined with Energy Dispersive Spectroscopy (EDS), Electron Backscattered Diffraction (EBSD) and Focused Ion Beam (FIB)/Transmission Electron Microscopy (TEM) with nano EDS. A particular attention was paid to the examination of the U(Mo) foil, the PVD coating, the cladding/Zr and Zr/U(Mo) interfaces. The Zr coating has a thickness around 15 ㎛. It has a columnar microstructure and appears dense. The cohesion of the cladding/Zr and Zr/U(Mo) interfaces seems to be satisfactory. An almost continuous layer with a thickness of the order of 100-300 nm is present at the cladding/Zr interface and corresponds to an oxidized part of the Zr coating. At the Zr/U(Mo) interface, a thin discontinuous layer is observed. It could correspond to locally oxidized U(Mo). This work provides a basis for interpreting the results of characterizations on EMPIrE irradiated plates.

• Nuclear Engineering and Technology
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• v.21 no.4
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• pp.332-340
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• 1989

Destructive examination of 14$\times$14 PWR fuel burned for 3 cycles are carried out to investigate the in-reactor fuel performance. The results obtained are as follows; 1) Grain growth is not occured at the fuel center. 2) Fuel density is decreased as the turnup increase, the density is down to 94.4% TD at burnup of 36,000 MWD/MTU. 3) Average thickness of oxide layer on cladding is less than 10 $\mu$m in the lower and middle section, while it is rapidly increased above 20 $\mu$m in the upper section. 4) The rate of hydride production in the cladding is large in the upper section than lower section and is related to the production of oxide on the cladding

• Nuclear Engineering and Technology
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• v.42 no.2
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• pp.193-202
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• 2010

In order to study the cladding properties of zirconium after a loss-of-coolant accident (LOCA)-simulation oxidation and water quenching test, commercial Zircaloy-4 and two kinds of HANA claddings were oxidized at temperatures ranging from $900^{\circ}C$ to $1250^{\circ}C$ and exposed for 300 s, and then cooled to $700^{\circ}C$ before quenching. Microstructural observations were made to evaluate the matrix characteristics with the chemical compositions after the LOCA-simulation test. Ring compression testing was then performed to compare the ductile behaviour of the HANA and Zircaloy-4 claddings. An X-ray diffraction (XRD) analysis was carried out for temperatures ranging from room temperature to $1250^{\circ}C$ for the oxide layer to verify the oxide crystal structure at each oxidation temperature.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3723-3740
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• 2021

The harsh conditions in the reactor affect the thermal and mechanical performance of the fuel plate heavily. Some in-pile behaviors, like fission-induced swelling, can cause a large deformation of fuel plate at very high burnup, which may even disturb the flow of coolant. In this research, the emphasis is put on the thermal expansion, fission-induced swelling, interaction layer (IL) growth, creep of the fuel meat, and plasticity of the cladding for the U₃Si₂/Al dispersion fuel plate. A detailed model of the fuel meat swelling is developed. Taking these in-pile behaviors into consideration, the three-dimensional large deformation incremental constitutive relations and stress update algorithms have been developed to study its thermal-mechanical performance under normal conditions using Abaqus. Results have shown that IL can effectively decrease the thermal conductivity of fuel meat. The high Mises stress region mainly locates at the interface between fuel meat and cladding, especially around the side edge of the interface. With irradiation time increasing, the stress in the fuel plate gets larger resulting from the growth of fuel meat swelling but then decreases under the effect of creep deformation. For the cladding, plasticity deformation does not occur within the irradiation time.

• Journal of Welding and Joining
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• v.13 no.2
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• pp.139-149
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• 1995

Many pressure vessels for the power plants are fabricated from low alloy ferritic steels. The inner sides of the pressure vessels are commonly weld_cladded with austenitic stainless steels to minimize problems of corrosive attack. The submerged-arc welding(SAW) process is now used in preference to other processes because of the possibilities open to automation to reduce the overaII welding times. The most reliable way to avoid underclad cracks(UCC) which are often detected at the overlap of the clad beads is to use nonsusceptible steels such as SA508 class 3. At present domestically developed forging steel of SA508 cl.S is now being cladded with single layer by using 90mm wide strip, which transfers higher heat input into the base metal compared to the conventional two layers strip cladding which has been in wide use with 30-60 mm wide strip. But the current indices for the influence of heat input on crack susceptibility are not accurate enough to express the subtle difference in crack susceptibility of the steel. Therefore, the purpose of this present study is: l) To determine UCC susceptibility on domestic forging steel, SA508 cl.S cladded with single layer by using submerged arc 90mm strip and, 2) To optimize heat input range by which the crack susceptibility could be eliminated.

• Wind and Structures
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• v.30 no.1
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• pp.99-117
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• 2020

This paper focuses on the processes of wind flow in atmospheric boundary layer, to produce realistic full scale pressures for design of low-rise buildings. CFD with LES turbulence closure is implemented on a scale 1:1 prototype building. A proximity study was executed computationally in CFD with LES that suggests new recommendations on the computational domain size, in front of a building model, apart from common RANS-based guidelines (e.g., COST and AIJ). Our findings suggest a location of the test building, different from existing guidelines, and the inflow boundary proximity influences pressure correlation and reproduction of peak loads. The CFD LES results are compared to corresponding pressures from open jet, full scale, wind tunnel, and the ASCE 7-10 standard for roof Component & Cladding design. The CFD LES shows its adequacy to produce peak pressures/loads on buildings, in agreement with field pressures, due to its capabilities of reproducing the spectral contents of the inflow at 1:1 scale.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3535-3542
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• 2023

Eutectic reactions of five kinds of Cr-coated Zr alloy cladding with different base materials (Zr-Nb-Sn alloy or Zr-Nb alloy), different coating thicknesses (6~22.5 mm), and different coating materials (Cr single layer or Cr/CrN bilayer) were studied using Differential Scanning Calorimetry (DSC). The DSC experiments demonstrated that the onset temperatures of the Cr single layer coated specimens were almost identical to ~1308 ℃, regardless of base materials or coating thicknesses. This study demonstrated that the Cr/CrN bilayer coated Zr-Nb-Sn alloy has a slightly (~10 ℃) higher eutectic onset temperature compared to the single Cr-coated specimen. The eutectic region characterized by post-eutectic microstructure proportionally increases with coating thickness. The post-eutectic characterization with different holding times at high temperature (1310-1330 ℃) reveals that progression of Zr-Cr eutectic requires time, and it dramatically changed with exposure time and temperature. The practical value of the time gain in non-instantaneous eutectic formation in terms of safety margin, however, seems to be limited.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.22 no.5
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• pp.554-560
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• 2021

CRUD (chalk river unidentified deposits) is a porous material deposited on the surface of nuclear fuel during nuclear power plant operation. The CRUD is composed of metal oxides, such as iron, nickel, and chromium. It is essential to investigate the effects of the CRUD layer on the wall heat transfer between the nuclear fuel surface and the coolant in the event of a nuclear accident. CRUD only negatively affects the temperature of the nuclear fuel due to heat resistance because the effects of the CRUD layer on two-phase boiling heat transfer are not considered. In this study, the physical property models for the porous CRUD layer were developed and implemented into the SPACE code. The effects of boiling heat transfer models on the peak cladding temperature and quenching were investigated by simulating a reflood experiment. The calculation results showed some positive effects of the CRUD layer.

• Journal of the Korean institute of surface engineering
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• v.35 no.6
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• pp.363-370
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• 2002

Empirical equations for the oxide thickness and the weight gain of Zircaloy-4 cladding during the oxidation in high temperature, high pressure steam have been developed. Firstly, the empirical equations for oxide thickness in 1 atm steam in 700~100$0^{\circ}C$ were made, then, the enhancement factor for the steam pressure effects on Zircaloy-4 cladding oxidation in high temperature steam was added. Based on the analysis of the weight fraction of dissolved oxygen in metal layer, empirical equations for the weight gain of Zircaloy-4 in high pressure, high temperature steam were developed. We compare the developed empirical equations with the Baker-Just correlation. The Baker-Just correlation can give a non-conservative estimation of oxidation of Zircaloy-4, depending on the steam pressure. These developed empirical equations can be used for the correct estimation of oxidation of Zircaloy-4 during accident analysis.

• Tribology and Lubricants
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• v.25 no.4
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• pp.250-255
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• 2009

Oxidation characteristics of Zirlo and Zircaloy-4 tubes, which are widely used as nuclear power fuel cladding, are studied in steam environment up to $1200^{\circ}C$. Oxidation resistances are compared in terms of the mass increase due to the absorption of oxygen. The evolution of microscopic structure accompanied with the oxidation process is investigated. Also, the influence of oxidation on the fretting wear characteristics of the tubes is studied. Piezo-electrically actuated rig is employed to fret the tubes with cross-contacting arrangement. Wear scar is observed and measured, by using microscopes and a 3D-profiler. The results of fretting wear are quantified in terms of scar size, wear volume and wear coefficient, and compared for the three different tube materials of oxidated Zirlo, virgin Zirlo and Zircaloy-4.