• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 추계학술발표회
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• pp.785-786
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• 2005

• 한국진공학회:학술대회논문집
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• 한국진공학회 2011년도 제40회 동계학술대회 초록집
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• pp.21-21
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• 2011

A series of Quasi-Elastic Neutron Scattering (QENS) experiments helps us to understand the single-particle (hydrogen atom) dynamics of a globular protein and its hydration water and strong coupling between them. We also performed Molecular Dynamics (MD) simulations on a realistic model of the hydrated hen-egg Lysozyme powder having two proteins in the periodic box. We found the existence of a Fragile-to-Strong dynamic Crossover (FSC) phenomenon in hydration water around a protein occurring at TL=$225{\pm}5K$ by analyzing Intermediate Scattering Function (ISF). On lowering of the temperature toward FSC, the structure of hydration water makes a transition from predominantly the High Density Liquid (HDL) form, a more fluid state, to predominantly the Low Density Liquid (LDL) form, a less fluid state, derived from the existence of a liquid?liquid critical point at an elevated pressure. We showed experimentally and confirmed theoretically that this sudden switch in the mobility of the hydration water around a protein triggers the dynamic transition (so-called glass transition) of the protein, at a temperature TD=220 K. Mean Square Displacement (MSD) is the important factor to show that the FSC is the key to the strong coupling between a protein and its hydration water by suggesting TL${\fallingdotseq}$TD. MD simulations with TIP4P force field for water were performed to understand hydration level dependency of the FSC temperature. We added water molecules to increase hydration level of the protein hydration water, from 0.30, 0.45, 0.60 and 1.00 (1.00 is the bulk water). These confirm the existence of the FSC and the hydration level dependence of the FSC temperature: FSC temperature is decreased upon increasing hydration level. We compared the hydration water around Lysozyme, B-DNA and RNA. Similarity among those suggests that the FSC and this coupling be universal for globular proteins, biopolymers.

• Nuclear Engineering and Technology
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• 제38권5호
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• pp.433-436
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• 2006

The implementation of multiplexing techniques combined with advances in neutron optics make the neutron three-axis spectrometers (TAS) an efficient tool to map inelastic response from single crystals over momentum transfer ranges comparable to the size of a single Brillouin zone. Thanks to recent progress in polarization techniques such experiments can be combined relatively easily with neutron polarization analysis, which does not only provide unambiguous separation of response corresponding to structural and magnetic degrees of freedom, but permits a quantitative analysis of the magnetic response anisotropy, often of crucial importance to test theoretical predictions. In the forthcoming decade we therefore expect a further development of the complementary use, rather than competition, of the reactor-based TAS's with time-of-flight (TOF) instruments for single crystal spectroscopy at the existing (ISIS) as well as at the newly built (SNS, J-PARK) pulsed sources.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3181-3204
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• 2022

The variational nodal method for solving the neutron transport equation has evolved over 40 years. Based on a functional form of the Boltzmann neutron transport equation, the method now comprises a complete set of variants that can be employed for different problems. This paper presents an extensive review of the development of the variational nodal method. The emphasis is on summarizing the whole theoretical system rather than validating the methodologies. The paper covers the variational nodal formulation of the Boltzmann neutron transport equation, the Ritz procedure for various application purposes, the derivation of boundary conditions, the extension for adjoint and perturbation calculations, and treatments for anisotropic scattering sources. Acceleration approaches for constructing response matrices and solving the resulting system of algebraic equations are also presented.

• 한국방사선학회논문지
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• 제13권7호
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• pp.917-924
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• 2019

양성자 치료기의 Passive Scattering System 노즐을 모의모사 하여 노즐 내 각 구성품에서 발생되는 중성자를 에너지별로 평가하였다. MCNPX code를 이용하여 치료환경에 사용되는 양성자 에너지 220 MeV, 도달거리 20 cm, 6 cm 길이의 SOBP를 구현하고, 치료기 가동 시 발생하는 중성자를 각 구성품에 따라 종류별로 분류하였다. 양성자 가속기 구성품 중 산란체에서 중성자가 가장 높게 발생되었으며 양성자의 중심 선속에서부터 멀어질수록 중성자의 선속은 감소되었다. 본 연구는 양성자 가속기의 유지 보수 및 해체에 필수적인 방사화 평가를 진행하기 위한 기초자료로 활용할 수 있을 것으로 사료된다.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.423-429
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• 2021

The investigation of the utilization of enriched ²⁰⁸Pb as a coolant to enhance the performance of a long-life fast reactor with a Modified CANDLE (Constant Axial shape of Neutron flux, nuclide densities, and power shape During Life of Energy production) burnup scheme has performed. The analyzes were performed on a reactor with thermal power of 800 MegaWatt Thermal (MWTh) with a refueling process every 15 years. Uranium Nitride (enriched ¹⁵N), ²⁰⁸Pb, and High-Cr martensitic steel HT-9 were employed as fuel, coolant, and cladding materials, respectively. One of the Pb-nat isotopes, ²⁰⁸Pb, has the smallest neutron capture cross-section (0.23 mb) among other liquid metal coolants. Furthermore, the neutron-producing cross-section (n, 2n) of ²⁰⁸Pb is larger than sodium (Na). On the other hand, the inelastic scattering energy threshold of ²⁰⁸Pb is the highest among Na, ^natPb, and Bi. The small inelastic scattering cross-section of ²⁰⁸Pb can harden the neutron energy spectrum. Therefore, ²⁰⁸Pb is a better neutron multiplier than any other liquid metal coolant. The excess neutrons cause more production than consumption of ²³⁹Pu. Hence, it can reduce the initial fuel loading of the reactor. The selective photoreaction process was developing to obtain enriched ²⁰⁸Pb. The neutronic was calculated using SRAC and JENDL 4.0 as a nuclear data library. We obtained that the modified CANDLE reactor with enriched ²⁰⁸Pb as coolant and reflector has the highest k-eff among all reactors. Meanwhile, the ^natPb cooled reactor has the lowest k-eff. Thus, the utilization of the enriched ²⁰⁸Pb as the coolant can reduce reactor initial fuel loading. Moreover, the enriched ²⁰⁸Pb-cooled reactor has the smallest power peaking factor among all reactors. Therefore, the enriched ²⁰⁸Pb can enhance the performance of a long-life Modified CANDLE fast reactor.

• Macromolecular Research
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• 제16권1호
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• pp.62-65
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• 2008

The morphology of the $D_2O$-induced reverse micellar structure of an amphiphilic block copolymer of poly( ethylene oxide )-b-poly(propylene oxide )-b-poly( ethylene oxide )($EO_{76}PO_{29}EO_{76}$) was investigated in hydrophobic media by small angle neutron scattering (SANS). Increasing $D_2O$ in the styrene/divinylbenzene solution of $EO_{76}PO_{29}EO_{76}$ led to a change in morphology of the reverse micelles from a short range ordered molecular aggregate to a hexagonally arranged micelle, and further to a spherical micelle.

• Journal of Radiation Protection and Research
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• 제14권1호
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• pp.66-70
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• 1989

방사성 중성자선원은 일상적 시험에 있어 표준 중성자 방사선장을 형성하는데 적합하다. 방사선 방어상의 목적으로 사용되는 중성자 측정기기의 교정을 위한 기준 방사선이 ISO TC-85에서 제의되었다. 한국표준연구소 방사선연구실에는 ISO TC-85의 추천사항에 준하여 개인용 중성자 선량계를 교정하기 위하여 $^{252}Cf$와 $^{241}Am-Be$ 선원을 이용한 표준조사시설을 설립하였다. 본 연구에서는 중성자 산란과 선원 비등방성에 연관된 교정상의 보정인자들을 실험에 의하여 결정하였다.

• Journal of Radiation Protection and Research
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• 제41권4호
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• pp.384-388
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• 2016

Background: For a verification of newly-developed neutron absorbers, one of guidelines on the qualification and acceptance of neutron absorbers is the neutron attenuation test. However, this approach can cause a problem for the qualifications that it cannot distinguish how the neutron attenuates from materials. Materials and Methods: In this study, an estimation method of neutron absorption performances for materials is proposed to detect both direct penetration and back-scattering neutrons. For the verification of the proposed method, MCNP simulations with the experimental system designed in this study were pursued using the polyethylene, iron, normal glass and the vitrified form. Results and Discussion: The results show that it can easily test neutron absorption ability using single absorber model. Also, from simulation results of single absorber and double absorbers model, it is verified that the proposed method can evaluate not only the direct thermal neutrons passing through materials, but also the scattered neutrons reflected to the materials. Therefore, the neutron absorption performances can be accurately estimated using the proposed method comparing with the conventional neutron attenuation test. Conclusion: It is expected that the proposed method can contribute to increase the reliability of the performance of neutron absorbers.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.989-995
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• 2018

Argonne National Laboratory of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have cooperated on the development, design, and construction of a neutron source facility. The facility was constructed at Kharkov, Ukraine, and its commissioning process is underway. The facility will be used for researches, producing medical isotopes, and training young nuclear specialists. The neutron source facility is designed with a provision to include a cryogenically cooled moderator system-a cold neutron source (CNS). This CNS provides low-energy neutrons, which will be used in the scattering experiment and material structures analysis. Cold neutron guides, coated with reflective material for the low-energy neutrons, will be used to transport the cold neutrons to the experimental site. The cold neutron guides would keep the cold neutrons within certain energy and angular space concentrated inside, while most of the gamma rays and high-energy neutrons are not affected by the cold neutron guides. For the KIPT design, the cold neutron guides need to extend several meters outside the main shield of the facility, and curved guides will also be used to remove the gamma and high-energy neutron. The neutron guides should be installed inside a shield structure to ensure an acceptable biological dose in the facility hall. Heavy concrete is the selected shielding material because of its acceptable performance and cost. Shield design analysis was carried out for the CNS guide hall. MCNPX was used as the major computation tool for the design analysis, with neutron and gamma dose calculated separately. Weight windows variance reduction technique was also used in the shield design. The goal of the shield design is to keep the total radiation dose below the $5.0{\mu}Sv/hr$ guideline outside the shield boundary. After a series of iterative MCNPX calculations, the shield configuration and parameters of CNS guide hall were determined and presented in this article.