• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 2005년도 봄학술 발표회 논문집(II)
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• pp.225-228
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• 2005

This paper was accomplished for analyzing the reason of the above deterioration happened on the deck of concrete bridge. The bridge was constructed at 660m above the sea level having more freezing and snowing days. Therefore, it is placed on the particular condition sprinkling $CaCl_2$ enough for keeping up with moderate traffic condition. When it is considered to the former condition, the bridge can be assumed to potentialities for combined deterioration with freezing-thawing under sprinkling deicing chemical. Core specimens were gathered from the concrete deck for clearing the reason of the above deterioration exactly, and it is used for various tests for measuring the compressive strength, elastic modulus, content of $Cl^-$, freezing-thawing at the fresh and salt water. As a result of freezing-thawing test, the specimen at the fresh water has over 90$\%$ of durability factor, but another specimen at 1$\%$ of salt water has 0$\%$ of durability factor at 140 cycles of the freezing-thawing. The result means that frost damage is sccelerated at the salt water. Therefore, the deterioration of the concrete deck is estimated to be occured by combined effects of freezing-thawing and chloride ion attack.

• 한국전기전자재료학회논문지
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• 제28권10호
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• pp.636-641
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• 2015

In this study, the partial discharge degradation properties for 2-core PVC cable($2cores{\times}1.5mm^2$ cross section, length of 10 cm, 20 cm, 30 cm) following immersion with the salt water that the 2%, 4%, 8% of NaCl is dissolved in 100 g of distilled water for 48 and 96 hours has been measured. The results of this study are as follows. When the degradation time in salt water of 2% NaCl is 48 hours, it found that the number of partial discharge increased as about 40 pps, 50 pps, 90 pps with increasing the length of cable to 10 cm, 20 cm, 30 cm. In case the concentration and degradation time is same, the inception and extinction voltage decreased with increasing the length of cable. When the degradation time in salt water is 96 hours and the length of cable is 20 cm, it found that the number of partial discharge decreased as 3,000 pps, 500 pps, 100 pps with increasing the concentration of NaCl to 2%, 4%, 8%.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.762-769
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• 2024

The TMSR-SF0 simulator is an integral effect thermal-hydraulic experimental system for the development of thorium molten salt reactor (TMSR) program in China. The simulator has two heat transport loops with liquid FLiNaK. In literature, the 95% level confidence uncertainties of the thermophysical properties of FLiNaK are recommended, and the uncertainties of density, heat capacity, thermal conductivity and viscosity are ±2%, ±10, ±10% and ±10% respectively. In order to investigate the effects of thermophysical properties uncertainties on the molten salt heat transport system, the uncertainty and sensitivity analysis of the heat transfer characteristics of the simulator system are carried out on a RELAP5 model. The uncertainties of thermophysical properties are incorporated in simulation model and the Monte Carlo sampling method is used to propagate the input uncertainties through the model. The simulation results indicate that the uncertainty propagated to core outlet temperature is about ±10 ℃ with a confidence level of 95% in a steady-state operation condition. The result should be noted in the design, operation and code validation of molten salt reactor. In addition, more experimental data is necessary for quantifying the uncertainty of thermophysical properties of molten salts.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.455-465
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• 2021

The Molten Salt Reactor (MSR), one of the six advanced reactor types of the 4th generation nuclear energy systems, has many impressive features including economic advantages, inherent safety and nuclear non-proliferation. This paper introduces a system analysis code named TREND, which is developed and used for the steady and transient simulation of MSRs. The TREND code calculates the distributions of pressure, velocity and temperature of single-phase flows by solving the conservation equations of mass, momentum and energy, along with a fluid state equation. Heat structures coupled with the fluid dynamics model is sufficient to meet the demands of modeling MSR system-level thermal-hydraulics. The core power is based on the point reactor neutron kinetics model calculated by the typical Runge-Kutta method. An incremental PID controller is inserted to adjust the operation behaviors. The verification and validation of the TREND code have been carried out in two aspects: detailed code-to-code comparison with established thermal-hydraulic system codes such as RELAP5, and validation with the experimental data from MSRE and the CIET facility (the University of California, Berkeley's Compact Integral Effects Test facility).The results indicate that TREND can be used in analyzing the transient behaviors of MSRs and will be improved by validating with more experimental results with the support of SINAP.

• Nuclear Engineering and Technology
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• 제56권12호
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• pp.5212-5221
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• 2024

This research focuses on the description and modeling of a one-dimensional molten salt reactor (MSR), in the presence of thermal feedback. Following the example of previous works, a simple one-dimensional system is proposed, describing a molten salt reactor with a main neutron-multiplying zone called core and a recirculation loop where the salt cools down. Specific attention is paid to the precursors' drift by modifying the neutron balance equation. Liquid nuclear fuels are characterized by a high volumetric expansion coefficient in comparison to customary solid fuels. Therefore, a strong coupling between neutronics and thermal-hydraulics is expected. As a consequence, a highly negative density coefficient characterizes the thermal feedback on the neutron reactivity. The precursor equation is here inverted analytically and combined with the neutron balance equation to obtain a generalized eigenvalue problem with the neutron flux distribution as the unknown. The balance equations are derived by finite volume integration over a discretized mesh, and the coupling between the two physical models is treated by Picard iterations. The numerical solution is finally extended to time-dependent calculations and compared to an analytical work for a one-dimensional circulating fuel reactor already existing in the literature.

• Elastomers and Composites
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• 제55권4호
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• pp.249-256
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• 2020

Superabsorbent polymers (SAPs) can absorb and retain ten to thousand times their dry mass of water because of their three-dimensional hydrophilic structures. Conventional SAPs are mainly composed of poly(acrylic acid sodium salt) derived from petrochemicals. The present work is aimed at limiting the use of the petrochemical component by replacing it with a biomass-based material. First, the core-SAP was prepared via the terpolymerization of itaconic acid, vinylsulfonic acid, and cellulose, and the optimum conditions in terms of material input ratio were determined. Following this, the core-SAP was surface-crosslinked by esterification with butane diol to improve its liquid permeability and absorbency under load (AUL). The liquid permeability was measured according to the amount of 0.9 wt.% NaCl solution passing between the swollen SAP particles under a given pressure, and the AUL was estimated from the weight of this solution absorbed under 0.3 psi pressure.

• 한국식물생명공학회:학술대회논문집
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• 한국식물생명공학회 2005년도 추계학술대회 및 한일 식물생명공학 심포지엄
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• pp.189-189
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• 2005

• 한국식물생명공학회:학술대회논문집
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• 한국식물생명공학회 2005년도 추계학술대회 및 한일 식물생명공학 심포지엄
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• pp.191-191
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• 2005

• 한국에너지공학회:학술대회논문집
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• 한국에너지공학회 1999년도 춘계 학술발표회 논문집
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• pp.113-118
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• 1999

AMBIDEXTER(Advanced Molten-salt Break-even Inherently-safe Dual-mission Experimental and TEst Reactor)는 고온저압의 Th/$^{233}$ U 불화용융염을 핵연료로 사용하므로 피복관이나 독립된 냉각재 없이 핵연료 자체가 열수송 매체로서 순환하는 원자로시스템개념으로서 저농축 $^{235}$ U 고체 핵연료를 사용하는 기존의 원자력 발전시스템이 안고있는 핵확산과 안전성 등의 고유문제를 해결할 수 있는 혁신형 차세대 원자력 발전시스템이다.(중략)

• 지질공학
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• 제28권1호
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• pp.35-45
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• 2018

이미징 기술은 매질 자체의 구조특성 뿐만 아니라 유체의 유동특성을 분석하기 위해 공극규모, 코어규모, 그리고 중급규모에서 다양하게 활용되고 있다. 본 기술보고에서는 코어규모에서 유동실험과 이미징기술을 연계하여 균질매질, 균열매질, 그리고 불균질매질에서의 이산화탄소($CO_2$) 유동 특성과 함께 $CO_2$ 주입시 주입관정 주변에서 발생할 수 있는 염침전 현상, 균질매질과 균열매질에서의 모세관압 평가, 그리고 $CO_2$ 주입이 완료된 이후 나타나는 포획메커니즘 평가 등에 대해 살펴보았다. 이미징기술을 통해 코어내에서 시간에 따른 $CO_2$ 플룸을 이미지화 함으로써 유동특성을 분석할 수 있으며, 특히 균열이 포함된 매질과 불균질한 매질에서의 유동 및 저장특성을 평가할 수 있다.