• Applied Microscopy
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• 제33권4호
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• pp.251-259
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• 2003

물질의 구조 특성파악에 많이 사용되는 X-선과 전자선에 대한 소스 원을 살펴보고 물질과의 반응을 atomic scattering factor의 항으로 설명하였다. 물질과의 회절을 역 격자 공간에서의 Ewald sphere로 설명하고 유한 크기의 소스 파장과 검출기의 효과도 함께 고려하였다.

• 한국결정학회지
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• 제12권1호
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• pp.25-30
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• 2001

• Nuclear Engineering and Technology
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• 제47권7호
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• pp.939-944
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• 2015

The calibration methods of neutron-measuring devices such as the neutron survey meter have advantages and disadvantages. To compare the calibration factors obtained by the shadow cone method and semi-empirical method, 10 neutron survey meters of five different types were used in this study. This experiment was performed at the Korea Atomic Energy Research Institute (KAERI; Daejeon, South Korea), and the calibration neutron fields were constructed using a $^{252}Californium$ ($^{252}Cf$) neutron source, which was positioned in the center of the neutron irradiation room. The neutron spectra of the calibration neutron fields were measured by a europium-activated lithium iodide scintillator in combination with KAERI's Bonner sphere system. When the shadow cone method was used, 10 single moderator-based survey meters exhibited a smaller calibration factor by as much as 3.1-9.3% than that of the semi-empirical method. This finding indicates that neutron survey meters underestimated the scattered neutrons and attenuated neutrons (i.e., the total scatter corrections). This underestimation of the calibration factor was attributed to the fact that single moderator-based survey meters have an under-ambient dose equivalent response in the thermal or thermal-dominant neutron field. As a result, when the shadow cone method is used for a single moderator-based survey meter, an additional correction and the International Organization for Standardization standard 8529-2 for room-scattered neutrons should be considered.

• Nuclear Engineering and Technology
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• 제34권5호
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• pp.482-493
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• 2002

The MOX fuel for LWR is fabricated either by direct mechanical blending of UO$_2$ and PuO$_2$ or by two stage mixing. Hence Pu-rich particles, whose Pu concentrations are higher than pellet average one and whose size distribution depends on a specific fabrication method, are inevitably dispersed in MOX pellet. Due to the inhomogeneous microstructure of MOX fuel, the thermal conductivity of LWR MOX fuel scatters from 80 to 100 % of UO$_2$ fuel. This paper describes a mechanistic thermal conductivity model for MOX fuel by considering this inhomogeneous microstructure and presents an explanation for the wide scattering of measured MOX fuel＇s thermal conductivity. The developed model has been incorporated into a KAERI＇s fuel performance code, COSMOS, and then evaluated using the measured in-pile data for MOX fuel. The database used for verification consists of homogeneous MOX fuel at beginning-of-life and inhomogeneous MOX fuel at high turnup. The COSMOS code predicts the thermal behavior of MOX fuel well except for the irradiation test accompanying substantial fission gas release. The over-prediction with substantial fission gas release seems to suggest the need for the introduction of a recovery factor to a term that considers the burnup effect on thermal conductivity.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.461-468
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• 2020

The improvements of the DeCART/MUSAD code system for uncertainty analysis of HTGR neutronic parameters are presented in this paper. The function for quantifying an uncertainty of critical-spectrumweighted few group cross section was implemented using the generalized adjoint B₁ equation solver. Though the changes between the infinite and critical spectra cause a considerable difference in the contribution by the graphite scattering cross section, it does not significantly affect the total uncertainty. To reduce the number of iterations of the generalized adjoint transport equation solver, the generalized adjoint B₁ solution was used as the initial value for it and the number of iterations decreased to 50%. To reflect the implicit uncertainty, the correction factor was derived with the resonance integral. Moreover, an additional correction factor for the double heterogeneity was derived with the effective cross section of the DH region and it reduces the difference from the complete uncertainty. The code system was examined with the MHTGR-350 Ex.II-2 3D core benchmark. The k_eff uncertainty for Ex.II-2a with only the fresh fuel block was similar to that of the block and the uncertainty for Ex.II-2b with the fresh fuel and the burnt fuel blocks was smaller than that of the fresh fuel block.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(1)
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• pp.185-190
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• 1996

RFSP is a computer program to do fuel management calculations for CANDU reactors. Its main function is to calculate neutron flux and power distributions using two-energy-group, three dimensional neutron diffusion theory. However, up to now the treatment has not been true two-group but actually "one-and-half groups". In other words, the previous (1.5-group) version of RFSP lumps the fast fission term into the thermal fission term. This is based on the POWDERPUFS-V Westcott convention. Also, there is no up-scattering term or bundle power over cell flux (H1 factor) for the fast group. While POWDERPUFS-V provides only 1.5 group properties, true two-group cross sections for the design and analysis of CAUDU reactors can be obtained from WIMS-AECL. To treat the full two-group properties, the previous RFSP version was modified by adding the fast fission, up-scatter terms, and H1 factor. This two-group version of RFSP is a convenient tool to accept lattice properties from any advanced lattice code (e.g. WIMS-AECL DRAGON, HELIOS...) and to apply to advanced fuel cycles. In this study, the modification to implement the true two-group treatment was performed only in the subroutines of the ＊SIMULATE module of RFSP. This module is the appropriate one to modify first, since it is used for the tracking of reactor operating histories. The modified two-group RFSP was evaluated with true two-group cross sections from WIMS-AECL. Some tests were performed to verify the modified two-group RFSP and to evaluate the effects of fast fission and up-scatter for three core conditions and four cases corresponding to each condition. The comparisons show that the two-group results are quite reasonable and serve as a verification of the modifications made to RFSP. To assess the long-term impact of the full 2-group treatment, it is necessary to simulate a long period (several months) of reactor history. It will also be necessary to implement the full two-group treatment of reactivity devices and assess the reactivity-device worths.ce worths.

• Nuclear Engineering and Technology
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• 제4권4호
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• pp.306-321
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• 1972

중성자산란을 이용하여 (NH$_4$)$_2$SO$_4$에 있어서 분자의 재배향(reorientation)을 조사하였다. T=300$^{\circ}$K의 측정에 대해서는 여러산란각도에서 분리된 준탄성산란스펙트럼과 구조인자(form-factor)를 SKOLD 이론에 의한 네개의 재배향모형과 비교하여 NH$_4$이온이 $\tau$$_{c}$=2.0$\times$$10^{-11}$ sec인 3중회전 4축(3-fold four axes) 또는 2중회전 3축재배향을 한다는 결론을 얻었다. $\tau$$_{c}$의 온도의존성을 100$^{\circ}$K-413$^{\circ}$K에 걸쳐 조사했으며, 고온상에 대해서는 복합스책트럼의 폭을 NMR이완시간측정에 얻은 결론과 비교하였다. 이상의 결과는 중성자산란이 고체에 있어서 분자재배향의 상세한 조사에 유력한 한 방법임을 보여 주었으며 따라서 이 방법의 응용에 대한 고찰을 하였다. 그 한 예로서 NH$_4$I(상 1)에 대해서 측정한 탄성구조인자와 자유회전근사에서 얻은 이론치를 비교하여 NH$_4$이온이 8면체형포텐샬 (Potential)에서 $\tau$$_{c}$$\leq$$10^{-12}$ sec인 재배향을 하고 있음을 주장하였다. 분자재배향이 비탄성 스펙트럼에 미치는 영향에 대하여 간단한 이론적 고찰을 하였다.을 하였다.

• Journal of Radiation Protection and Research
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• 제43권2호
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• pp.75-84
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• 2018

Background: Gamma-ray spectrometry helps in radiation shielding problems and different applications of radioisotopes. Experimental arrangements including broad beam geometries are widely used. The aim is to investigate and evaluate the ${\gamma}-ray$ spectra via attenuation by environmental materials. Materials and Methods: The photo peak to nominated parts in the ${\gamma}-ray$ spectra and the attenuation coefficients ${\mu}_b/{\rho}$ from broad beam geometries are measured for the materials water, soil, sand and cement at the energies 0.662, 1.25, and 1.332 MeV with a $3{^{\prime}^{\prime}}{\times}3{^{\prime}^{\prime}}$ NaI(Tl) detector. Results and Discussion: The ${\gamma}-ray$ spectra vary according to changes in the effective atomic number $Z_{eff}$ of the attenuator, the photon energy and the solid angle. The peak to total ratios are the most sensitive parts to variations in the experimental conditions and overturn in the region 0.663 MeV to 1.332 MeV. This is indicated as inversion trend. The results are discussed in view of $Z_{eff}$ and the experimental conditions. The intensity build-up is larger at the lower energy and larger scattering angles in agreement with Klein-Nishina formula and other results. The build-up factor B is$${\sim_=}$$1 at high ${\gamma}-energies$ and small scattering angles. Conclusion: The sensitivity to material characteristics decrease gradually from peak: to total, to Compton valley, to Compton plateau ratios. Rigorous collimation is necessary at small energies. Cement, of the largest $Z_{eff}$, is characterized by the maximum broad beam mass attenuation coefficients ${\mu}_b/{\rho}$. The obtained results provide information to decide for the suitable experimental set-up based on aim of the work.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2613-2620
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• 2023

Excellent thermal neutron absorption performance of boron expands the potential use of boron rich slag to prepare epoxy resin matrix nuclear shielding composites. However, shielding attenuation behaviors and mechanism of the composites against gamma rays are unclear. Based on the radiation protection theory, Phy-X/PSD, XCOM, and ⁶⁰Co gamma ray source were integrated to obtain the shielding parameters of boron rich slag/epoxy resin composites at 0.015-15 MeV, which include mass attenuation coefficient (µ_t), linear attenuation coefficient (µ), half value thickness layer (HVL), electron density (N_eff), effective atomic number (Z_eff), exposure buildup factor (EBF) and exposure absorption buildup factor (EABF).µ_t, µ, HVL, N_eff, Z_eff, EBF and EABF are 0.02-7 cm²/g, 0.04-17 cm^-1, 0.045-20 cm, 5-14, 3 × 10²³-8 × 10²³ electron/g, 0-2000, and 0-3500. Shielding performance is BS4, BS3, BS3, BS1 in descending order, but worse than ordinary concrete. µ and HVL of BS1-BS4 for ⁶⁰Co gamma ray is 0.095-0.110 cm^-1 and 6.3-7.2 cm. Shielding mechanism is main interactions for attenuation gamma ray by BS1-BS4 are elements with higher content or higher atomic number via Photoelectric Absorption at low energy range, and elements with higher content via Compton Scattering and Pair Production in Nuclear Field at middle and higher energy range.

• 한국전기전자재료학회논문지
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• 제15권1호
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• pp.67-74
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• 2002

A phosphor for Plasma Display Panel, BaMgAl$_{10}$ O$_{17}$ :Eu$^{2+}$, showing a blue emission band at about 450nm was prepared by a solid-state reaction using BaCO$_3$, $Al_2$O$_3$, MgO, Eu$_2$O$_3$ as starting materials wish flux AlF$_3$. The study of the behaviour of Eu in BAM phosphor was carried out by the photoluminescence spectra and the Rietveld method with X-ray and neutron powder diffraction data to refine the structural parameters such as lattice constants, the valence state of Eu, the preferential site of Mg atom and the site fraction of each atom. The phenomenon of the concentration quenching was abound 2.25~2.3wt% of Eu due to a decrease in the critical distance for energy transfer of inter-atomic Eu. Through the combined Rietveld refinement, R-factor, R$_{wp}$, was 8.11%, and the occupancy of Eu and Mg was 0.0882 and 0.526 at critical concentration. The critical distance of Eu$^{2+}$ in BAM was 18.8$\AA$ at 2.25% Eu of the concentration quenching. Furthermore, c/a ratio was decreased to 3.0wt% and no more change was observed over that concentration. The maximum entropy electron density was found that the modeling of $\beta$-alumina structure in BaMgAl$_{10}$ O$_{17}$ :Eu$^{2+}$correct coincided showing Ba, Eu, O atoms of z= 1/4 mirror plane.e.ane.e.