• Journal of Microbiology and Biotechnology
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• v.30 no.2
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• pp.216-225
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• 2020

An esterase gene, estA1, was cloned from Alteromonas sp. 39-G1 isolated from the Beaufort Sea. The gene is composed of 1,140 nucleotides and codes for a 41,190 Da protein containing 379 amino acids. As a result of a BLAST search, the protein sequence of esterase EstA1 was found to be identical to Alteromonas sp. esterase (GenBank: PHS53692). As far as we know, no research on this enzyme has yet been conducted. Phylogenetic analysis showed that esterase EstA1 was a member of the bacterial lipolytic enzyme family IV (hormone sensitive lipases). Two deletion mutants (Δ20 and Δ54) of the esterase EstA1 were produced in Escherichia coli BL21 (DE3) cells with part of the N-terminal of the protein removed and His-tag attached to the C-terminal. These enzymes exhibited the highest activity toward p-nitrophenyl (pNP) acetate (C₂) and had little or no activity towards pNP-esters with acyl chains longer than C₆. Their optimum temperature and pH of the catalytic activity were 45℃ and pH 8.0, respectively. As the NaCl concentration increased, their enzyme activities continued to increase and the highest enzyme activities were measured in 5 M NaCl. These enzymes were found to be stable for up to 8 h in the concentration of 3-5 M NaCl. Moreover, they have been found to be stable for various metal ions, detergents and organic solvents. These salt-tolerant and chemical-resistant properties suggest that the enzyme esterase EstA1 is both academically and industrially useful.

• The Journal of Korean Academy of Prosthodontics
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• v.50 no.3
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• pp.149-155
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• 2012

Purpose: The purpose of this study was to investigate the influence of mobilization on bone-implant interface prior to osseointegration of fixtures. Materials and methods: The experimental implants (3.75 mm in diameter, 4.0 mm in length) were made of commercially pure (Grade IV) titanium, and were treated with RBM ($MegaGen^{(R)}$: Ca-P). The 80 implants (two in each tibia) were inserted into the monocortical tibias of 20 rabbits which each weighed more than 3.5 kg (Female, New Zealand White). According to the removal torque interval, the groups were divided into 10 groups, Group I (6 wks), Group II (4 days+6 wks), Group III (4 days+1 wk+6 wks), Group IV (1 wk+6 wks), Group V (1 wk+1 wk+6 wks), Group VI (2 wks+6 wks), Group VII (2 wks+ 1 wk+6 wk), Group VIII (3 wks+6 wks), Group IX (3 wks+1 wk+6 wks) and Group X (10 wks). The control groups were Group I and X, the removal torque was measured at 6 wks and 10 wks with a digital torque gauge (Mark-10, USA). In the experimental groups, the removal torque was given once or twice before the final removal torque and the value was measured each time. After which, the implants were put back where they had been except the control groups. All the experimental groups were given a final healing time (6 wks) before the final removal torque test, in which values were compared with the control groups and the 1st and/or 2nd removal torque values in each experimental group. Results: In the final removal torque tests, the removal torque value of Group X (10 wks) was higher than that of Group I (6 wks) in the control groups but not statistically different. There were no significant differences between the experimental groups and control groups (P>.05). In the first removal torque comparison, the experimental groups (4 days or 1 wk) values were significantly lower than the other experimental groups (2 wks or 3 wks). In the comparison of each experimental group according to healing time, the final removal torque value was significantly higher than the 1st torque test value. Conclusion: Once or twice mobilization of fixture prior to osseointegration did not deter the final bone to implant osseointegration, if sufficient healing time was given.

• Nuclear Engineering and Technology
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• v.38 no.1
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• pp.45-60
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• 2006

Presently, the VHTGR (Very High Temperature Gas-cooled Reactor) is considered the most attractive candidate for a GEN-IV reactor to produce hydrogen, which will be a key resource for future energy production. A new concept for a reactor cavity cooling system (RCCS), a critical safety feature in the VHTGR, is proposed in the present study. The proposed RCCS consists of passive water pool and active air cooling systems. These are employed to overcome the poor cooling capability of the air-cooled RCCS and the complex cavity structures of the water-cooled RCCS. In order to estimate the licensibility of the proposed design, its performance and integrity were tested experimentally with a reduced-scale mock-up facility, as well as with a separate-effect test facility (SET) for the 1/4 water pool of the RCCS-SNU to examine the heat transfer and pressure drop and code capability. This paper presents the test results for SET and validation of MARS-GCR, a system code for the safety analysis of a HTGR. In addition, CFX5.7, a computational fluid dynamics code, was also used for the code-to-code benchmark of MARS-GCR. From the present experimental and numerical studies, the efficacy of MARS-GCR in application to determining the optimal design of complicated systems such as a RCCS and evaluation of their feasibility has been validated.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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• pp.101-106
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• 2016

NSSS (Nuclear Steam Supply System) and BOP (Balance of Plant) design works for PGSFR (Prototype Gen-IV Sodium Fast Reactor) have been conducted in Korea. NSSS major components, e.g. reactor vessel, steam generator and secondary sodium main pipes, are designed according to the rule of ASME boiler and pressure vessel code division 5, in which DBA (Design by Analysis) methods are used in the stress assessments. However, there is little discussions about detail rules for BOP piping design. In this paper, the detail methodologies of BOP piping stress assessment are discussed including safety systems and non-safety system pipings. It is confirmed that KEPIC MGE(ASME B31.1) and ASME BPV code division 5 HCB-3600 can be used in stress assessments of non-safety pipes and class B pipes, respectively. However, class A pipe design according to ASME BPV code division 5 HBB-3200 has many difficulties applying to PGSFR BOP design. Finally, future development plan for class A pipe stress assessment method is proposed in this paper.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1131-1136
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• 2020

Xenon behaves differently in molten salt reactors (MSRs) compared to solid fuel reactors. This behavior needs exploring due to the large reactivity effect of the ¹³⁵Xe isotope, given the current interest in MSR power plant development for commercial deployment. This paper focuses on select topics in xenon transport, reviews relevant past works, and proposes specific research questions to advance the state of the art in each of the focus areas. Specifically, the paper discusses the issue of xenon solubility in MSRs, the behavior of particulates circulating in MSR fuel salt and its influence on the xenon transport, the possibility of ionization of xenon atoms which changes its effective size and thus affects its mass transport, and finally the issue of circulating void fraction and how it is measured. This work presents specific recommendations for MSR designers to research the limits of Henry's law validity, circulating particulate scrubbers, validity of mass transport coefficients in high radiation fields, and the effects of pump speed on circulating void fraction.

• Journal of computational fluids engineering
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• v.21 no.1
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• pp.19-29
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• 2016

A Prototype Gen-IV Sodium-cooled Fast Reactor which is one of the $4^{th}$ generation nuclear reactors is in development by Korea Atomic Energy Research Institute. The reactor is composed of four main fluid systems which are categorized by its functions, i.e., Primary Heat Transport System, Intermediate Heat Transport System, Decay Heat Removal System and Sodium-Water Reaction Pressure Relief System. The coolant of the reactor is liquid sodium and sodium-to-sodium heat exchangers are installed at the interfaces between two fluid systems, Intermediate Heat Exchangers between the Primary Heat Transport System and the Intermediate Heat Transport System and Decay Heat Exchangers between the Primary Heat Transport System and the Decay Heat Removal System. For the design and performance analysis of the Intermediate Heat Exchanger and the Decay Heat Exchanger, a computer code was written during previous step of research. In this work, the computer code named "SHXSA" has been validated preliminarily by computational fluid dynamics simulations.

• Nuclear Engineering and Technology
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• v.41 no.1
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• pp.1-10
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• 2009

The safe and reliable performance of fusion and fission plants depends on the choice of suitable materials and an assessment of long-term materials degradation. These materials are degraded by their exposure to extreme conditions; it is necessary, therefore, to address the issue of long-term damage evolution of materials under service exposure in advanced plants. The empirical approach to the study of structural materials and fuels is reaching its limit when used to define and extrapolate new materials, new environments, or new operating conditions due to a lack of knowledge of the basic principles and mechanisms present. Materials designed for future Gen IV systems require significant innovation for the new environments that the materials will be exposed to. Thus, it is a challenge to understand the materials more precisely and to go far beyond the current empirical design methodology. Breakthrough technology is being achieved with the incorporation in design codes of a fundamental understanding of the properties of materials. This paper discusses the multi-scale, multi-code computations and multi-dimensional modelling undertaken to understand the mechanical properties of these materials. Such an approach is envisaged to probe beyond currently possible approaches to become a predictive tool in estimating the mechanical properties and lifetimes of materials.

• Nuclear Engineering and Technology
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• v.51 no.3
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• pp.654-664
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• 2019

The geometrical and electromagnetic variables of a rectangular-type magnetohydrodynamic (MHD) circulation system are optimized to solve MHD equations for the active decay heat removal system of a prototype Gen-IV sodium fast reactor. Decay heat must be actively removed from the reactor coolant to prevent the reactor system from exceeding its temperature limit. A rectangular-type MHD circulation system is adopted to remove this heat via an active system that produces developed pressure through the Lorentz force of the circulating sodium. Thus, the rectangular-type MHD circulation system for a circulating loop is modeled with the following specifications: a developed pressure of 2 kPa and flow rate of $0.02m^3/s$ at a temperature of 499 K. The MHD equations, which consist of momentum and Maxwell's equations, are solved to find the minimum input current satisfying the nominal developed pressure and flow rate according to the change of variables including the magnetic flux density and geometrical variables. The optimization shows that the rectangular-type MHD circulation system requires a current of 3976 A and a magnetic flux density of 0.037 T under the conditions of the active decay heat removal system.

• Nuclear Engineering and Technology
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• v.51 no.1
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• pp.116-124
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• 2019

$M^2LFR-1000$ is a medium-power modular lead-cooled fast reactor, developed by University of Science and Technology of China (USTC), aiming at achieving a reactor design fulfilling the Gen IV nuclear system requirements and meanwhile emphasizing the optimum safety and economics. In order to evaluate the safety performance of $M^2LFR-1000$ reactor core, three typical transients are selected from initiating events, which are unprotected transient overpower (UTOP), unprotected loss of offsite power (ULOHS+ULOF) and increase of feedwater flowrate with primary pumps trip (IFW+PLOF). These three transients presented and discussed in this paper are performed with the code Analysis of THermal-hydraulics of LEaks and Transients (ATHLET), which is developed by Gesellschaft $f{\ddot{u}}r$ Anlagen-und Reaktorsicherheit gGmbH (GRS). The results indicate that the $M^2LFR$ is safe enough under these three transients due to the good inherent safety features of the reactor, without human intervention, the reactor will reach a new steady state under UTOP condition.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4125-4133
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• 2022

KAERI (Korea Atomic Energy Research Institute) developed the conceptual design of PGSFR (Prototype Gen-IV Sodium Cooled Fast Reactor) and Burner Reactor. Since the reactor characteristics of the PGSFR and Burner Reactor are different, the shape, size and the arrangement of the main components in the reactors must be different. Therefore, the conceptual design for the fuel handling and transfer systems needs to be performed coinciding with the structure of the reactor. Especially, because a redan structure dividing hot and cold pool is installed in the reactor vessel, the conceptual design of the fuel handling and transfer system largely changes depending on the location of the redan structure. Various elements of the conceptual design and an integral arrangement for the fuel handling and transfer system were arranged according to the characteristics, sizes and shapes of the reactors. In this paper, the conceptual designs of the fuel handling and transfer system for PGSFR and Burner Reactor are described. Especially, an A-frame method is selected as the fuel handling and transfer system for the Burner Reactor, considering the layout of the internal structure. The tilt angle, diameter and length of A-frame is determined and the strength evaluation of the A-frame is performed.