• Journal of Magnetics
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• v.18 no.3
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• pp.342-347
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• 2013

Local wall thinning in pipelines affects the structural integrity of industries, such as nuclear power plants (NPPs). In the present study, a development of pulsed eddy current (PEC) technology that detects the wall thinning of pipelines covered with insulation is reviewed. The methods and experimental results, which have two kinds of probe with a single and double core, were compared. For this purpose, the single and double core probes having one and two excitation coils have been devised, and the differential probe with two Hall sensors has been fabricated to measure the wall thinning in insulated pipelines. The test sample is a stainless steel having different thickness, laminated by plastic insulation to simulate the pipelines in NPPs. The excitation coils in the probe is driven by a rectangular current pulse, the difference of two Hall sensors has been measured as a resultant PEC signal. The peak value of the detected signal is used to describe the wall thinning. The double core probe has better performance to detect the wall thinning covered with insulation; the single core probe can detect the wall thinning up to an insulation thickness of 18 mm, whereas the double probe can detect up to 25 mm. The results show that the double core PEC probe has the potential to detect the wall thinning in an insulated pipeline of the NPPs.

• Journal of Powder Materials
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• v.12 no.4 s.51
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• pp.266-272
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• 2005

Transition metal ions($Ni^{2+}$, $Cr^{3+}$ and $V^{5+}$) doped $TiO_2$ nanostructured powders were synthesized by mechanical alloying(MA) to shift the adsorption threshold into the visible light region. The synthesized powders were characterized by XRD, SEM, TEM and BET for structural analysis, UV-Vis and photoluminescence spectrum for the optical study. Also, photocatalytic abilities were evaluated by decomposition of 4-chlorophenol(4CP) under ultraviolet and visible light irradiations. Optical studies showed that the absorption wavelength of transition metal ions doped $TiO_2$ powders moved to visible light range, which was believed to be induced by the energy level change due to the doping. Among the prepared $TiO_2$ powders, $NiO^{2+}$ doped $TiO_2$ powders, showed excellent photooxidative ability in 4CP decomposition.

• Nuclear Engineering and Technology
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• v.32 no.2
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• pp.99-107
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• 2000

A probabilistic integrity analysis method is presented for a reactor vessel under pressurized thermal shock(PTS) based on Monte Carlo simulation. This method can be applied to the structural integrity assessment of a reactor vessel subjected to pressurized thermal shock where the coolant temperature transient cannot be expressed explicitly as a time function. An axially or circumferentially oriented infinite length surface crack is assumed to be in the beltline weld region of the rector vessel's inside surface. The random variables are the initial crack depth, neutron fluence on the vessel's inside surface, the copper and nickel content of the vessel materials, R $T_{NDT}$ , $K_{IC}$ , and K/aub la/. The reliability of a sample reactor vessel under PTS is assessed quantitatively and the influence of the amount of neutron fluence is also examined by applying the present method.sent method.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.2966-2976
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• 2023

A tritium breeder is a lithium-based material capable of producing tritium. Many researchers designing nuclear-fusion energy are studying tritium production using pebbles, which are solid-type breeders. The sphericity and size of the pebbles are critical in obtaining pebbles with good tritium-breeding efficiency. Furthermore, tritium-release efficiency can be increased by using pebbles with appropriate porosities. Promising raw materials for tritium-breeding materials include Li₄SiO₄ and Li₂TiO₃. Li₄SiO₄ has a higher lithium density than Li₂TiO₃ and exhibits excellent tritium-breeding efficiency. However, it has the disadvantage of being easily decomposed during the Li₄SiO₄-green-pebble sintering process because of its low structural stability at high temperatures and high lithium density. In this study, we attempted to determine the optimal conditions for manufacturing Li₄SiO₄ pebbles using the droplet-freeze-drying method. The optimal Li₄SiO₄ slurry conditions and sintering temperatures were determined. The optimal Li₄SiO₄ slurry-fabrication conditions were 3 wt% polyvinyl alcohol and 75 wt% Li₄SiO₄ based on the deionized-water weight content. The sintering temperature at which Li₄SiO₄ did not decompose and exhibited the optimum porosity of 10.8% was 900 ℃.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.3
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• pp.297-306
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• 2022

The acoustic emission (AE) is proposed as a feasible method for the real-time monitoring of the structural damage evolution in concrete materials that are typically used in the storage of nuclear wastes. However, the characteristics of AE signals emitted from concrete structures subjected to various environmental conditions are poorly identified. Therefore, this study examines the AE characteristics of the concrete structures during uniaxial compression, where the storage temperature and immersion conditions of the concrete specimens varied from 15℃ to 75℃ and from completely dry to water-immersion, respectively. Compared with the dry specimens, the water-immersed specimens exhibited significantly reduced uniaxial compressive strengths by approximately 26%, total AE energy by approximately 90%, and max RA value by approximately 70%. As the treatment temperature increased, the strength and AE parameters, such as AE count, AE energy, and RA value, of the dry specimens increased; however, the temperature effect was only minimal for the immersed specimens. This study suggests that the AE technique can capture the mechanical damage evolution of concrete materials, but their AE characteristics can vary with respect to the storage conditions.

• Nuclear Engineering and Technology
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• v.52 no.4
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• pp.791-796
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• 2020

Nuclear fusion seems to be a good choice of energy source in the future. Nickel is one of the crucial structural materials for fusion devices. In this work, the cross section data of ⁵⁸Ni(n,p)⁵⁸Co, ⁶⁰Ni(n,p)⁶⁰Co, ⁶¹Ni(n,p)⁶¹Co, ⁶²Ni(n,p)⁶²Co and ⁶⁴Ni(n,p)⁶⁴Co reactions were calculated using the nuclear codes ALICE/ASH, EMPIRE 3.2 and TALYS 1.8. In addition, the cross sections were calculated with the empirical formulas obtained in our previous paper at 14-15 MeV. The obtained results were compared with the measured values in the literature, and with the evaluated data files (JEFF-3.3, TENDL-2017, ENDF/B-VIII.0).

• Nuclear Engineering and Technology
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• v.41 no.1
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• pp.11-20
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• 2009

Radiation effects on materials are inherently multiscale phenomena in view of the fact that various processes spanning a broad range of time and length scales are involved. A multiscale modeling approach to embrittlement of pressure vessel steels is presented here. The approach includes an investigation of the mechanisms of defect accumulation, microstructure evolution and the corresponding effects on mechanical properties. An understanding of these phenomena is required to predict the behavior of structural materials under irradiation. We used molecular dynamics (MD) simulations at an atomic scale to study the evolution of high-energy displacement cascade reactions. The MD simulations yield quantitative information on primary damage. Using a database of displacement cascades generated by the MD simulations, we can estimate the accumulation of defects over diffusional length and time scales by applying kinetic Monte Carlo simulations. The evolution of the local microstructure under irradiation is responsible for changes in the physical and mechanical properties of materials. Mechanical property changes in irradiated materials are modeled by dislocation dynamics simulations, which simulate a collective motion of dislocations that interact with the defects. In this paper, we present a multi scale modeling methodology that describes reactor pressure vessel embrittlement in a light water reactor environment.

• Smart Structures and Systems
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• v.24 no.6
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• pp.759-768
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• 2019

In nuclear fusion reactors, the key structural component (i.e., the plasma-facing component) undergoes high heat flux cyclic loading. To ensure the safety of fusion reactors, an experimental study on the temperature-induced creep of stainless steel under heat flux cyclic loading was performed in the present work. The strains were measured using a stereo digital image correlation technique (3D-DIC). The influence of the heat haze was eliminated, owing to the use of a vacuum environment. The specimen underwent heat flux cycles ($500^{\circ}C-1000^{\circ}C$) with different mechanical preloads (0 kN, 10 kN, 30 kN, and 50 kN). The results revealed that, for a relatively large preload (for example, 50 kN), a single temperature cycle can induce a residual strain of up to $15000{\mu}{\varepsilon}$.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1593-1601
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• 2021

Oxide dispersion strengthening (ODS) tungsten alloys are highly desirable in irradiation applications. However, how to improve the properties of ODS-tungsten alloys efficiently has been worth studying for a long time. Here we report a nanostructuring approach that achieves W-La₂O₃ alloy with a high level of flexural strength and Vickers hardness at room temperature, which have the maximum value of 581 MPa and 703 Hv, respectively. This method named solution combustion synthesis (SCS) can generate 30 nm coating structures W-La₂O₃ composite powders by using Keggin-type structural polyoxometalates as raw materials in a fast and low-cost process. The composite powder can be fabricated to W-La₂O₃ alloy with an optimal microstructure of submicrometric W grains coexisting with nanometric oxide particles in the grain interior, and a stability interface structure of grain boundaries (GBs) by forming transition zones. The method can be used to prepare new ODS alloys with excellent properties in the future.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1264-1272
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• 2016

The accurate estimation of leak rate through cracks is crucial in applying the leak before break (LBB) concept to pipeline design in nuclear power plants. Because of its importance, several programs were developed based on the several proposed flow models, and used in nuclear power industries. As the flow models were developed for a homogeneous pipe material, however, some difficulties were encountered in estimating leak rates for bimaterial pipes. In this paper, a flow model is proposed to estimate leak rate in bimaterial pipes based on the modified Henry-Fauske flow model. In the new flow model, different crack morphology parameters can be considered in two parts of a flow path. In addition, based on the proposed flow model, a program was developed to estimate leak rate for a crack with linearly varying cross-sectional area. Using the program, leak rates were calculated for through-thickness cracks with constant or linearly varying cross-sectional areas in a bimaterial pipe. The leak rate results were then compared and discussed in comparison with the results for a homogeneous pipe. The effects of the crack morphology parameters and the variation in cross-sectional area on the leak rate were examined and discussed.