• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.785-786
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• 2005

• Proceedings of the Korean Vacuum Society Conference
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• 2011.02a
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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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• v.38 no.5
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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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• v.54 no.9
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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.

• Journal of the Korean Society of Radiology
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• v.13 no.7
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• pp.917-924
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• 2019

The passive scattering system nozzle of the proton therapy accelerator was simulated to evaluate the neutrons generated by each component in each nozzle by energy. The Monte Carlo N-Particle code was used to implement spread out Bragg peak with proton energy 220 MeV, reach 20 cm, and 6 cm length used in the treatment environment. Among the proton accelerator components, neutrons were the highest in scatterers, and the neutron flux decreased as it moved away from the central flux of the proton. This study can be used as a basic data for the evaluation of the radiation necessary for the maintenance and dismantling of proton accelerators.

• Nuclear Engineering and Technology
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• v.53 no.2
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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.

• Journal of Radiation Protection and Research
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• v.14 no.1
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• pp.66-70
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• 1989

In routine testing, the radioactive neutron sources are particularly suitable for producing standard. neutron fields. The ISO TC-85 has proposed neutron reference radiation for the calibration of neutron measuring devices used for radiation protection purposes. Radiation laboratory of KSRI has installed a standard irradiation facility using $^{252}Cf$ and $^{241}Am-Be$ sources for calibrating personal dosimeters according to the recommendations given in ISO TC-85. In this study, correction factors for calibration related to neutron scattering and anisotropy are obtained by experiments with commercial rem meter for demonstration purposes.

• Macromolecular Research
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• v.16 no.1
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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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• v.41 no.4
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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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• v.50 no.6
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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.