• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.8 no.2
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• pp.1-6
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• 2012

A PHE (Process Heat Exchanger) in a nuclear hydrogen system is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR (Very High Temperature gas cooled Reactor) to the chemical reaction that yields a large quantity of hydrogen. A small-scale PHE prototype made of Hastelloy-X is being tested in a small-scale gas loop at Korea Atomic Energy Research Institute. Previous research on the high-temperature structural analysis of the small-scale PHE prototype had been performed only using parent material properties. In this study, high-temperature structural analysis using weld properties in weld zone was performed and the analysis results compared with the previous research.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.7 no.3
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• pp.48-53
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• 2011

A PHE (Process Heat Exchanger) is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR (Very High Temperature Reactor) to the chemical reaction that yields a large quantity of hydrogen. A small-scale PHE prototype made of Hastelloy-X is being tested in a small-scale gas loop at Korea Atomic Energy Research Institute. In this study, as a part of the evaluation on the high-temperature structural integrity of the small-scale PHE prototype, we carried out macroscopic high-temperature structural analysis of the small-scale PHE prototype under the gas loop test conditions considering the pipeline stiffness.

• Nuclear Engineering and Technology
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• v.47 no.6
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• pp.678-699
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• 2015

Although the harsh space environment imposes many severe challenges to space pioneers, space exploration is a realistic and profitable goal for long-term humanity survival. One of the viable and promising options to overcome the harsh environment of space is nuclear propulsion. Particularly, the Nuclear Thermal Rocket (NTR) is a leading candidate for nearterm human missions to Mars and beyond due to its relatively high thrust and efficiency. Traditional NTR designs use typically high power reactors with fast or epithermal neutron spectrums to simplify core design and to maximize thrust. In parallel there are a series of new NTR designs with lower thrust and higher efficiency, designed to enhance mission versatility and safety through the use of redundant engines (when used in a clustered engine arrangement) for future commercialization. This paper proposes a new NTR design of the second design philosophy, Korea Advanced NUclear Thermal Engine Rocket (KANUTER), for future space applications. The KANUTER consists of an Extremely High Temperature Gas cooled Reactor (EHTGR) utilizing hydrogen propellant, a propulsion system, and an optional electricity generation system to provide propulsion as well as electricity generation. The innovatively small engine has the characteristics of high efficiency, being compact and lightweight, and bimodal capability. The notable characteristics result from the moderated EHTGR design, uniquely utilizing the integrated fuel element with an ultra heat-resistant carbide fuel, an efficient metal hydride moderator, protectively cooling channels and an individual pressure tube in an all-in-one package. The EHTGR can be bimodally operated in a propulsion mode of $100MW_{th}$ and an electricity generation mode of $100MW_{th}$, equipped with a dynamic energy conversion system. To investigate the design features of the new reactor and to estimate referential engine performance, a preliminary design study in terms of neutronics and thermohydraulics was carried out. The result indicates that the innovative design has great potential for high propellant efficiency and thrust-to-weight of engine ratio, compared with the existing NTR designs. However, the build-up of fission products in fuel has a significant impact on the bimodal operation of the moderated reactor such as xenon-induced dead time. This issue can be overcome by building in excess reactivity and control margin for the reactor design.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.1
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• pp.1-8
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• 2009

Nuclear hydrogen production using high temperature heat of a very high temperature reactor(VHTR) is one of the most attractive ways of mass hydrogen production without greenhouse gas emission. In many countries, sulfur-iodine(S-I) thermochemical process and high temperature steam electrolysis(HTSE) process are being investigated. In such processes, corrosion behavior of Intermediate heat exchanger materials are the most critical issues. Especially in a HTSE system, several heat exchangers will be facing hot steam conditions. In this paper, the status of high temperature corrosion researches in hot steam and supercritical water conditions are reviewed in view of the implication to HTSE conditions. Based on the review, test condition and plan of the hot steam corrosion of the candidate materials are formulated and described in some details along with the schematics of the test set-up. The test results and subsequent evaluation will be used in development of a interface system between the HTSE hydrogen production system and the VHTR.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3990-4002
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• 2021

CSPACE (Core meltdown, Safety and Performance Analysis CodE for nuclear power plants) for a simulation of severe accident progression in a Pressurized Water Reactor (PWR) is developed by coupling of verified system thermal hydraulic code of SPACE (Safety and Performance Analysis CodE for nuclear power plants) and core damage progression code of COMPASS (Core Meltdown Progression Accident Simulation Software). SPACE is responsible for the description of fluid state in nuclear system nodes, while COMPASS is responsible for the prediction of thermal and mechanical responses of core fuels and reactor vessel heat structures. New heat transfer models to each phase of the fluid, flow blockage, corium behavior in the lower head are added to COMPASS. Then, an interface module for the data transfer between two codes was developed to enable coupling. An implicit coupling scheme of wall heat transfer was applied to prevent fluid temperature oscillation. To validate the performance of newly developed code CSPACE, we analyzed typical severe accident scenarios for OPR1000 (Optimized Power Reactor 1000), which were initiated from large break loss of coolant accident, small break loss of coolant accident, and station black out accident. The results including thermal hydraulic behavior of RCS, core damage progression, hydrogen generation, corium behavior in the lower head, reactor vessel failure were reasonable and consistent. We demonstrate that CSPACE provides a good platform for the prediction of severe accident progression by detailed review of analysis results and a qualitative comparison with the results of previous MELCOR analysis.

• Journal of Radiation Protection and Research
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• v.22 no.2
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• pp.97-101
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• 1997

During the period of March 1995 to December 1995, tritium concentrations of tritiated water vapor (HTO) and tritiated hydrogen (HT) in the atmosphere in Taejon were measured to evaluate present background levels of tritium in the atmosphere. Air samples were collected continuously for three weeks with a sampling system for tritium in the atmosphere and were analyzed by a liquid scintillation counting system. The range of the atmospheric HTO concentrations was 3.2-36 mBq $m^{-3}$ with a mean value of 16.2 mBq $m^{-3}$. The atmospheric HTO concentrations were the highest in summer and the lowest in winter. This trend was similar to the variation of atmospheric absolute humidity. The specific activities of tritium in atmospheric water vapor in Taejon ranged from 0.62 Bq $L^{-1}$ to 3.82 Bq $L^{-1}$ with a mean value of 2.04 Bq $L^{-1}$. The atmospheric HT concentrations were in the range of 35.7 mBq $m^{-3}$ to 48.9 mBq $m^{-3}$ with a mean value of 41.1 mBq $m^{-3}$.

• Nuclear Engineering and Technology
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• v.35 no.3
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• pp.226-233
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• 2003

The discharge characteristics of a prototype ion source was investigated, which was developed and upgraded for the NBI (Neutral Beam Injection) heating system of KSTAR (Korea Superconducting Tokamak Advanced Research). The ion source was designed for the arc discharge of magnetic bucket chamber with multi-pole cusp fields. The ion source was discharged by the emission-limited mode with the control of filament heating voltage. The maximum ion density was 4 times larger than the previous discharge controlled by a space-charge-limited mode with fully heated filament. The plasma (ion) density and arc current were proportional to the filament voltage, but the discharge efficiency was inversely proportional to the operating pressure of hydrogen gas. The maximum ion density and arc current were obtained with constant arc voltage ($80{\sim}100V$), as $8{\times}10^{11}cm^{-3}$ and 1200 A, respectively. The estimated maximum beam current was about 35 A, extracted by the accelerating voltage of 80kV.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3961-3968
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• 2024

The Korea Atomic Energy Research Institute (KAERI) is developing a novel Two-Region Arc Plasma Ion Source (TRAP) as a negative hydrogen (deuterium) ion source for a Neutral Beam Injection (NBI) system in a fusion tokamak. The TRAP ion source is based on a two-region configuration, comprising a high energy electron region that creates highly vibrationally excited molecules and a low electron temperature region that generates negative ions by attaching electrons to molecules. This configuration can be achieved by optimizing the filament position and magnetic cusp field. In order to optimize the TRAP configuration, the plasma parameters are investigated under various operating conditions, such as filament position, gas pressure, and arc power. Electron density and temperature are determined using Langmuir probe measurements. In this paper, the detailed experimental results are described and discussed.

• Biomolecules & Therapeutics
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• v.22 no.2
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• pp.136-142
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• 2014

We investigated the protective effects of chlorogenic acid (CGA), a polyphenol compound, on oxidative damage induced by UVB exposure on human HaCaT cells. In a cell-free system, CGA scavenged 1,1-diphenyl-2-picrylhydrazyl radicals, superoxide anions, hydroxyl radicals, and intracellular reactive oxygen species (ROS) generated by hydrogen peroxide and ultraviolet B (UVB). Furthermore, CGA absorbed electromagnetic radiation in the UVB range (280-320 nm). UVB exposure resulted in damage to cellular DNA, as demonstrated in a comet assay; pre-treatment of cells with CGA prior to UVB irradiation prevented DNA damage and increased cell viability. Furthermore, CGA pre-treatment prevented or ameliorated apoptosis-related changes in UVB-exposed cells, including the formation of apoptotic bodies, disruption of mitochondrial membrane potential, and alterations in the levels of the apoptosis-related proteins Bcl-2, Bax, and caspase-3. Our findings suggest that CGA protects cells from oxidative stress induced by UVB radiation.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.63-70
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• 2023

Ultrasound-assisted dissolution of U₃O₈ powder in carbonate solution was explored to determine if and how ultrasound act during the dissolution. The variation of U₃O₈ solid particles and uranyl complexes under ultrasound treatment and magnetic stirring was observed in carbonate media. The results show that the use of ultrasound can increase the solubility and dissolution rate of U₃O₈ powder than that under magnetic stirring. The crush of U₃O₈ particles and the reduction of the activation energy (Ea, kJ/mol) of U₃O₈ dissolution reaction were observed, which both play an important role in the ultrasonic-assisted dissolution of U₃O₈ in carbonate-peroxide solution. Meanwhile, there is no observation of the ultrasound effect on the distribution of uranyl species and hydrolysis of uranyl complexes during the ultrasound treatment in carbonate-peroxide solution. Although the generation of ·OH radicals under ultrasound (22 ± 2 kHz) was observed, the oxidation of ·OH had little effect on the dissolution of U₃O₈ in the carbonate-peroxide solution system.