• Journal of radiological science and technology
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• v.37 no.3
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• pp.223-231
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• 2014

Medical linear accelerator is widely used in radiation treatment field, and high energy photons, above 10 MV nominal accelerator voltage, are commonly utilized for the radiation treatment. However, these high energy photons lead the photo-nuclear reaction and the generation of photo-neutrons are accompanied. Thus, these problematic factors are issued in the view of radiation protection. Therefore, linear accelerator and radiation treatment room are simulated from MCNPX program in this study. The measurement points of interest are selected and analyzed, and the resulting effects derived from the properties of photo-neutron are evaluated. Therefore, we realized that the number of generating photo-neutrons was decreased by depending on the distance from the source. No matter what the nominal energy is set, the rates thermal neutrons to fast neutrons are marginal. It is founded that the amount of the thermal neutrons were decreased by depending on the distance from the source.

• Proceedings of the Korean Institute of Electrical and Electronic Material Engineers Conference
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• 2002.07a
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• pp.210-213
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• 2002

The blue phosphor, $BaMgAl_{10}O_{17}$:$Eu^{2+}$, showing a blue emission band at about 450 nm were prepared by solid state reaction of BaC $O_3$, A $l_2$ $O_3$, MgO and E $u_2$ $O_3$ with Al $F_3$ as a flux. The thermal quenching of BaMgAl $O_{17}$:E $u^{2+}$ phosphor significantly reduces the intensity of the blue emission. It is reduced by an amount of 50% after heating at around 800$^{\circ}C$ for 1 hr. The red emission in the 580∼720 nm region of $^{5}$ $D_{0}$\longrightarro $w^{7}$ $F_1$ and $^{5}$ $D_{0}$\longrightarro $w^{7}$ $F_2$ transition of $Eu^{3+}$ is produced from the phosphor heated above 1,100$^{\circ}C$. The EPR spectrum also reveals that some part of E $u^{2+}$ ions are oxidized to trivalent ions above 1,100$^{\circ}C$ at around 90 and 140mT. This oxidation evidence is also detected from XANES absorption spectra for $L_{III}$ shell of Eu ions: an absorption peak is at 6,977eV of E $u^{2+}$ and 6,984eV of $Eu^{3+}$. The combined X-ray and neutron data suggests that the new phase of EuMgA $l_{11}$ $O_{19}$ magnetoplumbite structure may be formed by heat treatment.eat treatment.tment.eat treatment.tment.t.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.230-237
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• 2020

In this part, an implicit time dependent solution is presented for the Boltzmann transport equation discretized by the analytic coarse mesh finite difference method (ACMFD) over the spatial domain as well as the simplified P₃ (SP₃) for the angular variable. In the first part of this work we proposed a SP₃-ACMFD approach to solve the static eigenvalue equations which provide the initial conditions for temp dependent equations. Having solved the 3D multi-group SP₃-ACMFD static equations, an implicit approach is resorted to ensure stability of time steps. An exponential behavior is assumed in transverse integrated equations to establish a relationship between flux moments and currents. Also, analytic integration is benefited for the time-dependent solution of precursor concentration equations. Finally, a multi-channel one-phase thermal hydraulic model is coupled to the proposed methodology. Transient equations are then solved at each step using the GMRES technique. To show the sufficiency of proposed transient SP₃-ACMFD approximation for a full core analysis, a comparison is made using transport peers as the reference. To further demonstrate superiority, results are compared with a 3D multi-group transient diffusion solver developed as a byproduct of this work. Outcomes confirm that the idea can be considered as an economic interim approach which is superior to the diffusion approximation, and comparable with transport in results.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.11
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• pp.2219-2227
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• 2002

In a nuclear power plant, reactor pressure vessel (RPV) is the primary pressure boundary component that must be protected against failure. The neutron irradiation on RPV in the beltline region, however, tends to cause localized damage accumulation, leading to crack initiation and propagation which raises RPV integrity issues. The objective of this paper is to estimate the integrity of RPV under hot leg leaking accident by applying the finite element analysis. In this paper, a parametric study was performed for various crack configurations based on 3-dimensional finite element models. The crack configuration, the crack orientation, the crack aspect ratio and the clad thickness were considered in the parametric study. The effect of these parameters on the maximum allowable nil-ductility transition reference temperature ($(RT_{NDT})$) was investigated on the basis of finite element analyses.

• Proceedings of the Materials Research Society of Korea Conference
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• 2003.03a
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• pp.112-112
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• 2003

최근 자성반도체(diluted magnetic semiconductor; DMS)를 이용한 소자 개발이 가긍해짐에 따라 국내외에서 활발한 연구가 이루어지고 있다. 본 연구실에서는 GaN-단일전구체를 이용하여 상온에서 자기적 특성을 나타내는 p-type GaMnN를 성장시켰다 극한 환경에서의 자성반도체 재료의 물성 변화를 알아보기 위해, 본 연구에서는 세계 최초로 중성자 빔의 조사에 따른 자성반도체의 구조적, 자기적 특성 및 열처리에 따른 특성 변화를 관찰 및 분석하였다. Molecular beam epitaxy(MBE)를 이용하여 Mn cell 온도가 각각 77$0^{\circ}C$, 94$0^{\circ}C$인 GaMnN 박막을 성장시켰다. 성장된 박막 시편에 한국원자력연구소 하나로 HTS공에서 중성자 빔을 각각 20min(4.17$\times$$10^{16}$n/$\textrm{cm}^2$), 24hour(3.0$\times$$10^{18}$n/$\textrm{cm}^2$)씩 조사하였다 중성자 빔을 조사한 시편은 진공분위기 하에서 100$0^{\circ}C$, 30초간 열처리하였다.(rapid thermal annealing;RTA, 승온속도: 8$^{\circ}C$/sec) 중성자 빔을 조사한 GaMnN 박막의 구조적인 특성은 X-ray diffraction(XRD) 측정을 통해 관찰하였고, 박막의 자기적 특성은 superconducting quantum interference device(SQUID)를 통해 측정하였다.

• Carbon letters
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• v.7 no.3
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• pp.196-200
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• 2006

Graphite has hexagonal closed packing structure with two bonding characteristics of van der Waals bonding between the carbon layers at c axis, and covalent bonding in the carbon layer at a and b axis. Graphite has high tolerant to the extreme conditions of high temperature and neutron irradiations rather than any other materials of metals and ceramics. However, carbon elements easily react with oxygen at as low as 400C. Considering the increasing production of today of hydrogen and electricity with a nuclear reactor, study of oxidation characteristics of graphite is very important, and essential for the life evaluation and design of the nuclear reactor. Since the oxidation behaviors of graphite are dependent on the shapes of testing specimen, critical care is required for evaluation of nuclear reactor graphite materials. In this work, oxidation rate and amounts of the isotropic graphite (IG-110, Toyo Carbon), currently being used for the Koran nuclear reactor, are investigated at various temperature. Oxidation process or principle of graphite was figured out by measuring the oxidation rate, and relation between oxidation rate and sample shape are understood. In the oxidation process, shape effect of volume, surface area, and surface to volume ratio are investigated at $600^{\circ}C$, based on the sample of ASTM C 1179-91.

• Nuclear Engineering and Technology
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• v.31 no.4
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• pp.391-400
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• 1999

A conceptual core design of the 1,300MWe KNGR (Korean Next Generation Reactor) without using soluble boron for reactivity control was developed to determine whether it is technically feasible to implement SBF (Soluble Boron Free) operation. Based on the borated KNGR core design, the fuel assembly and control rod configuration were modified for extensive use of burnable poison rods and control rods. A new fuel rod, in which Pu-238 had been substituted for a small amount of U-238 in fuel composition, was introduced to assist the reactivity control by burnable poison rods. Since Pu-238 has a considerably large thermal neutron capture cross section, the new fuel assembly showed good reactivity suppression capability throughout the entire cycle turnup, especially at BOC (Beginning of Cycle). Moreover, relatively uniform control of power distribution was possible since the new fuel assemblies were loaded throughout the core. In this study, core excess reactivity was limited to 2.0 %$\delta$$\rho$ for the minimal use of control rods. The analysis results of the SBF KNGR core showed that axial power distribution control can be achieved by using the simplest zoning scheme of the fuel assembly Furthermore, the sufficient shutdown margin and the stability against axial xenon oscillations were secured in this SBF core. It is, therefore, concluded that a SBF operation is technically feasible for a large sized LWR (Light Water Reactor).

• Analytical Science and Technology
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• v.18 no.4
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• pp.263-270
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• 2005

Nuclear track is an useful tool for elemental analysis of radionuclides, such as uranium, plutonium and thorium, etc., and for elements undergoing nuclear reactions with thermal neutrons such as lithium and boron. This method has various application fields such as detecting fissionable radionuelides, measuring the fission rate in nuclear technology, analyzing cosmic radiation from meteorite, calculating the age of minerals as well as their history, etc. Track registration method has been applied to the microscopic analysis of boron and fissionable element such as uranium in KAERI. This report reviews the theoretical background of the nuclear track formation, practical procedures to obtain etched tracks and a perspective of the future.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.61 no.12
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• pp.1948-1952
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• 2012

The paper presents modal testing and analysis in order to obtain the dynamic characteristics of a middle spacer grids of a nuclear fuel rod. A spacer grid is one of the important structural elements supporting nuclear fuel rods. Such a fuel rod can be oscillated by its thermal expansion, neutron irradiation and etc. due to cooling water flow under the operation of a nuclear power plant. When the fuel rod vibrates, fretting wear due to repeated friction motion between the fuel rods and spacer grids can be occurred, and so the fuel rod is damaged. In this paper, through modal analysis and testing, natural frequencies and modes of a middle spacer grid were calculated, and the following conclusions were obtained. Firstly the numerical first-seven natural frequencies for spacer grids of a fuel rod having complicated structures have a small difference within 3.8% with experimental natural frequencies, and so the suitability of simulation results was verified. Secondly, experimental mode shapes for a middle spacer grid of a nuclear fuel rod were verified by obtaining lower non-diagonal terms through MAC(Modal Assurance Criteria), and were confirmed by the simulation modes.

• Corrosion Science and Technology
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• v.2 no.6
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• pp.253-259
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• 2003

Zirconium based alloys have been extensively used as a cladding material for fuel rods in nuclear reactors, due to their low thermal neutron absorption cross-section, excellent corrosion resistance and good mechanical properties at high temperatures. However, a cladding material for fuel rods in nuclear reactors was contact water during long time at high-temperature, so it is necessary to improve the wear and corrosion resistance of the fuel cladding, At ambient environment, there are few data or paper on the characteristic of corrosion in chloride solution and acidic solution. The specimens used in this work are pure Zr and Zircaloy-4. Zircaloy-4 is a specific zirconium-based alloy containing, on a weight percent basis, 1.4% Sn, 0.2% Fe, 0.1% Cr. Pitting corrosion resistance of two alloys by ASTM G48 is higher than that of electrochemical method. Passive film formed on Zircaloy-4 is mainly composed of $ZrO_2$, metallic Sn, and iron species regardless of formation environments. Also, passive film formed on Zr alloys shows n-type semiconductic property on the base of Mott-Schottky plot.