• Nuclear Engineering and Technology
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• 제55권1호
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• pp.144-155
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• 2023

The core catcher is used as a passive safety system in new generation nuclear power plants to create a space in the containment for the placing and cooling of the molten corium under various severe accidents. This research investigates the role of the core catcher in the VVER-1000 reactor containment system in mitigating the effects of core meltdown under various severe accidents within the context of the Ex-vessel Melt Retention (EVMR) strategy. Hence, a comparison study of three severe accidents is conducted, including Station Black-Out (SBO), SBO combined with the Large Break Loss of Coolant Accident (LB-LOCA), and SBO combined with the Small Break Loss of Coolant Accident (SB-LOCA). Numerical comparative simulations are performed for the aforementioned scenario with and without the EX-vessel core-catcher. The results showed that considering the EX-Vessel core catcher reduces the amount of hydrogen by about 18.2 percent in the case of SBO + LB-LOCA, and hydrogen production decreases by 12.4 percent in the case of SBO + SB-LOCA. Furthermore, in the presence of an EX-Vessel core-catcher, the production of gases such as CO and CO2 for the SBO accident is negligible. It was revealed that the greatest decrease in pressure and temperature of the containment is related to the SBO accident.

• 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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• 제56권3호
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• pp.793-802
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• 2024

In the framework of the European project SAMOSAFER, this numerical study focuses on some thermal aspects of the Emergency Draining Tank (EDT) located underneath the core of a Molten Salt Reactor. In case of an emergency, this tank passively receives the liquid fuel salt and is designed to ensure a subcritical state. An important requirement is that the fuel does not overheat to maintain the EDT Hastelloy container integrity. The present EDT is based upon a group of hexagonal cooling assemblies arranged in a hexagonal grid and cooled down thanks to conduction through the inert salt layer up to an air flow in charge of removing the heat. This numerical thermal study relies on a conjugated heat transfer analysis coupling a Finite Element solid thermal code (SYRTHES) and two instances of a Finite Volume CFD codes (Code_Saturne). Calculations on an initial design suggest that a simple center airpipe flow is likely to not sufficiently cool the device. Alternative solutions have been evaluated. Introduction of fins to enhance the heat transfer do not bring a noticeable improvement regarding maximum temperature reached. However, a solution in which the central pipe air flow is replaced by several cooling channels located closer to the fuel is investigated and suggests a better cooling.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1996년도 춘계학술발표회논문집(2)
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• pp.669-674
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• 1996

The analysis for the coolability of the reactor cavity in typical Korean 1000 MWe Nuclear Unit under severe accidents is performed using MELCOR 1.8.3 code. The key parameters molten core-concrete interaction(MCCI) such as melt temperature, concrete ablation history and gas generation are investigated. Total twenty cases are selected according to ejected debris fraction and coolant mass, The ablation rate of concrete decreases as mass of the melt decreases and coolant mass increases. Heat loss from molten pool to coolant is comparable to total decay heat, so concrete ablation is delayed until water is absent and crust begins to remove. Also, overpressurization due to non-condensible gases generated during corium and concrete interacts can cause to additional risk of containment failure. It is concluded that flooded reactor cavity condition is very important to minimize the cavity ablation and pressure load by non-condensible gases on containment.

• 한국분말야금학회:학술대회논문집
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• 한국분말야금학회 2006년도 Extended Abstracts of 2006 POWDER METALLURGY World Congress Part 1
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• pp.604-605
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• 2006

Recently the Marangoni convention is supposed to be an important phenomenon that significantly affects the solidification. For understanding the Marangoni convection mechanism, visualizing the convention phenomenon of molten tin with ultrasonic has been conducted. This paper reports developing a tracer material of micro metal balloon that is used in the molten system. We have succeeded in coating the surface of Shirasu-ballons with nickel by plating process. The obtained metal balloon is spherical and some characterizations were conducted.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4551-4559
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• 2022

The aim of this research is to analyze the characteristics of a core melt discharged from the reactor vessel and the spreading behavior the core melt in the reactor cavity of the SMART. First, a severe accident sequence under conservative conditions is simulated by the MELCOR code to obtain the conditions for an analysis of the spreading behavior and coolability of the ex-vessel melt. Second, the spreading behavior and coolability of the ex-vessel melt are analyzed by the MELTSPREAD code. The level, temperature, and pressure of the water in the cavity as well as the temperature, mass, composition, and discharge velocity of the melt were utilized to construct the ex-vessel analysis. The melt spread only to part of the cavity, and that the height of the corium in a static state was less than 25 cm. The characteristics of a small modular reactor on the spreading behavior and coolability of melt were analyzed. In the SMART, the amount of melt discharged into the cavity is relatively small and the area of the cavity is sufficiently large when compared to a high-power pressurized water reactor. It was found that the coolability of an ex-vessel core melt can be sufficiently secured.

• Journal of Advanced Marine Engineering and Technology
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• 제30권5호
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• pp.629-635
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• 2006

Manufacture of motor by laser has been studying realization that was demands on market for lightening and miniaturization. Moreover. early in the 1980s. manufacture of parts for automobiles by laser welding was already successfully introduced. The purpose of this study was to develop production technology of the high quality laminated stator core for motor by pulsed Nd:YAG laser heat source. In the event of adjusting defocus and voltage to control humping in laser welding of the laminated core. sound bead could be obtained. but deep penetration was not. Therefore. explosive evaporating plasma was controlled by adjustment of peak power on pulse width. Particularly, because explosive evaporating plasma induced high peak power, made molten metal in keyhole scatter. a suitable adjustment of peak power was required to obtain sound bead. As a results of experiment. sound bead and deep penetration could be obtained.

• 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)

• 대한전기학회:학술대회논문집
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• 대한전기학회 1998년도 추계학술대회 논문집 학회본부 C
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• pp.861-863
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• 1998

As-casted high nitrogen alloys (Fe-25%Mn-12%Cr-1%Ni) have been investigated to study core material. Nitrogen concentration in molten alloys was increased with increasing the square root of nitrogen gas pressure in melting chamber. This result can be explained by Sievert's law. Nitrogen that dissolved as a interstital solid solution element in austenite stainless steel increased lattice parameter and hardness. Electric resistivity($\rho$) was increased with increasing nitrogen concentration and was about $80{\mu}{\Omega}cm$ at room temperature. Coefficient of linear thermal expansion of the nitrogen steel was about $22{\times}10^{-6}/^{\circ}C$.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.973-979
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• 2024

In the framework of the Generation IV research and development project, in which the French Commission of Alternative and Atomic Energies (CEA) is involved, a main objective for the design of Sodium-cooled Fast Reactor (SFR) is to meet the safety goals for severe accidents. Among the severe ones, the Unprotected Transient OverPower (UTOP) accidents can lead very quickly to a global melting of the core. UTOP accidents can be considered either as slow during a Control Rod Withdrawal (CRW) or as fast. The paper focuses on fast UTOP accidents, which occur in a few milliseconds, and three different scenarios are considered: rupture of the core support plate, uncontrolled passage of a gas bubble inside the core and core mechanical distortion such as a core flowering/compaction during an earthquake. Several levels and rates of reactivity insertions are also considered and the thermal-mechanical behavior of an ASTRID fuel pin from the ASTRID CFV core is simulated with the GERMINAL code. Two types of fuel pins are simulated, inner and outer core pins, and three different burn-up are considered. Moreover, the feedback from the CABRI programs on these type of transients is used in order to evaluate the failure mechanism in terms of kinetics of energy injection and fuel melting. The CABRI experiments complete the analysis made with GERMINAL calculations and have shown that three dominant mechanisms can be considered as responsible for pin failure or onset of pin degradation during ULOF/UTOP accident: molten cavity pressure loading, fuel-cladding mechanical interaction (FCMI) and fuel break-up. The study is one of the first step in fast UTOP accidents modelling with GERMINAL and it has shown that the code can already succeed in modelling these type of scenarios up to the sodium boiling point. The modeling of the radial propagation of the melting front, validated by comparison with CABRI tests, is already very efficient.