• 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.

• 기술혁신학회지
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• 제6권4호
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• pp.462-479
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• 2003

The policy trends of use and the development of nuclear power in the United States, France, United Kingdom. Germany, Russia, China, Japan and Korea are briefly investigated. Nuclear power technology has been developed as the national policy in the nuclear-advanced countries. 50 years has passed since the declaration of "Atoms for Peace" by USA President Eisenhour in December 1953. Recently, it appears to revitalize the nuclear power program in world major countries in order to recover the shortage of electric power and to curb the excess emission of carbon dioxide as well as to secure competitiveness in electricity markets. Advanced countries are making new initiatives for the development of the fourth generation nuclear power system. Furthermore, wide-ranged use and development of nuclear power technologies are expected in district heating in commercial sectors, power in the space exploration, and propulsion power of large tankers and spaceships. High temperature gas cooled nuclear power reactor will be applied for mass production of hydrogen energy in the future.

• Nuclear Engineering and Technology
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• 제17권1호
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• pp.8-15
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• 1985

Debris bed내·외로부터 깨스유량을 갖는 debris/water 열적상호작용 해석모델이 중대사고 분석을 위해 제시되었다. 제시된 모델은 증기 소비, debris bed에서 수소 생성, 유입깨스 및 화학반응열에 대한 인자들을 포함하고 있으며, 금속-물반응 및 debris/concrete 작용으로 인한 깨스 생성을 평가하기 위해 MARCH code에 도입시켰다. 그 결과 수소원은 격납용기 과도압력에 큰 영향을 미치나 debris bed로 대류깨스 냉각과 콘크리트로 전도 열손실은 debris bed 냉각성에 조그마한 영향을 주는 것으로 나타났다. 하지만 debris 인자의 재가열과 재용융은 콘크리트와 상호작용에 의해 상당히 지연될 수 있다.

• 한국수소및신에너지학회논문집
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• 제22권4호
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• pp.546-551
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• 2011

The corrosion on steam generator tubes of the secondary side of pressurized water reactor inhibits heat transfer. One of the most efficient techniques improving the heat transfer performance of a nuclear electric generation is a corrosion control. The environmental parameters mostly affecting corrosion are materials and chemical additives. It seems that no further corrosion occurs in steels with Polyacrylic acid polymer dispersant treatment. Polyacrylic acid forms a protective coating with uniform thickness on metal surface. Polyacrylic treatment appears to be the most convenient way to enhance the thermal performance by the thermal conductivity improvement in steam generators.

• 한국광학회지
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• 제29권3호
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• pp.119-125
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• 2018

수소 가스는 신 재생 에너지원으로서, 에너지의 발생과정에서 오염물질의 배출이 없는 친환경적인 에너지원이다. 그러나 수소 가스는 인화 에너지가 낮으며, 무색, 무취의 화염 전파성과 폭발성이 강한 매우 위험한 물질 중 하나이다. 수소 가스의 검출을 위한 다양한 기술이 있으나, 대부분 센서를 이용하여 대기 중의 수소 가스를 수집하여 측정하는 근거리 측정 기술이 대부분이다. 수소 가스 측정 기술 중 하나인 라만 라이다 장치는 수소 가스의 강한 라만 산란 현상을 이용하여 원거리에서 수소 가스 농도 검출 및 분포를 계측할 수 있는 장점을 갖고 있다. 본 연구에서는 넓은 영역에서(2~50 m) 누출된 수소 가스의 원거리 측정을 위한 on-axis 형태를 갖는 라만 라이다 장치를 개발하였다. 본 연구를 통하여 개발된 수소 가스 원거리 탐지 거리가 향상된 라만 라이다 장치의 성능을 검증하기 위하여, 수소 가스 폭발을 방지 및 농도 변화가 가능한 가스 챔버를 이용하여 라만 라이다 장치로부터 50 m 거리에 위치한 수소 가스 농도 측정 실험을 수행하였다. 그 결과, 개발된 라만 라이다 장치를 이용하여 50 m 거리에 위치한 0.66 Vol.%의 수소 가스 검출이 가능함을 증명하였다.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2670-2677
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• 2023

In the nuclear safety studies, a new trend refers to the evaluation of uncertainties as a mandatory component of best-estimate safety analysis which is a modern and technically consistent approach being known as BEPU (Best Estimate Plus Uncertainty). The major objectives of this study consist in performing a study of uncertainties/sensitivities of the major analysis results for a generic CANDU 6 Nuclear Power Plant during Station Blackout (SBO) progression to understand and characterize the sources of uncertainties and their effects on the key figure-of-merits (FOMs) predictions in severe accidents (SA). The FOMs of interest are hydrogen mass generation and event timings such as the first fuel channel failure time, beginning of the core disassembly time, core collapse time and calandria vessel failure time. The outcomes of the study, will allow an improvement of capabilities and expertise to perform uncertainty and sensitivity analysis with severe accident codes for CANDU 6 Nuclear Power Plant.

• 대한기계학회논문집A
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• 제33권10호
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• pp.1108-1118
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• 2009

A Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source of the order of $950^{\circ}C$ for nuclear hydrogen generation, which can produce hydrogen from water or natural gas. A primary hot gas duct (HGD) as a coaxial double-tube type cross vessel is a key component connecting a reactor pressure vessel and an intermediate heat exchanger in the VHTR. In this study, a structural sizing methodology for the primary HGD of the VHTR is suggested in order to modulate a flow-induced vibration (FIV). And as an example, a structural sizing of the horizontal HGD with a coaxial double-tube structure was carried out using the suggested method. These activities include a decision of the geometric dimensions, a selection of the material, and an evaluation of the strength of the coaxial double-tube type cross vessel components. Also in order to compare the FIV characteristics of the proposed design cases, a fluid-structure interaction (FSI) analysis was carried out using the ADINA code.

• Nuclear Engineering and Technology
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• 제45권6호
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• pp.767-790
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• 2013

MAAP4 is a computer code that can simulate the response of a light water reactor power plant during severe accident sequences, including actions taken as part of accident management. The code quantitatively predicts the evolution of a severe accident starting from full power conditions given a set of system faults and initiating events through events such as core melt, reactor vessel failure, and containment failure. Furthermore, models are included in the code to represent the actions that could mitigate the accident by in-vessel cooling, external cooling of the reactor pressure vessel, or cooling the debris in containment. A key element tied to using a code like MAAP4 is an uncertainty analysis. The purpose of this paper is to present a MAAP4 based analysis to examine the sensitivity of a key parameter, in this case hydrogen production, to a set of model parameters that are related to a Level 2 PRA analysis. The Level 2 analysis examines those sequences that result in core melting and subsequent reactor pressure vessel failure and its impact on the containment. This paper identifies individual contributors and MAAP4 model parameters that statistically influence hydrogen production. Hydrogen generation was chosen because of its direct relationship to oxidation. With greater oxidation, more heat is added to the core region and relocation (core slump) should occur faster. This, in theory, would lead to shorter failure times and subsequent "hotter" debris pool on the containment floor.

• Toxicological Research
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• 제17권4호
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• pp.249-253
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• 2001

Fumonisin B1, a mycotoxin, is thought to induce esophageal cancer in humans and apoptosis in animal cells by inhibiting ceramide synthase. Dumonisin Bl may also generate reactive oxygen species directly or indirectly, leading to DNA damage and lipid peroxidation. In this study, a DNA fragmentation assay, dichlorofluorescein (DCF) analysis, and single cell gel electrophoresis (SCGE) were used to investigate the involvement of cellular free radicals, specifically hydrogen peroxide, in the DNA damaging activity of fumonisin B1. From an in vitro DNA fragmentation assay, E. coli DNA, damage by fumonisin Bl was increased by the addition of superxide dismutase (SOD) and decreased by catalase. SCGE and DCF analysis in vivo showed that the nuclear DNA damage and intracellular free radicals in cultured rat hepatocytes treated with fumonisin B1 were increased with the concentration of fumonisin Bl . DNA damage and free radical generation were inhibited by the addition of catalase. Fumonisin Bl , in the presence of SOD, produces hydrogen peroxide causing oxidative DNA damage and protein malfunction, leading to genotoxicity and cytotoxicity of the toxin.

• 한국수소및신에너지학회논문집
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• 제22권1호
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• pp.118-127
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• 2011

An IGCC was evaluated as one of the next generation technologies that would be able to substitute for coal-fired power plants. According to "The 4th Basic Plan of Long-term Electricity Supply & Demand" which is developed by the Electricity Business Acts, the first IGCC will be operated at 2015. Like other new and renewable energy such as solar PV, Fuel cell, The IGCC is considered as non-competitive generation technology because it is not maturity technology. Before the commercial operation of an IGCC in our electricity market, its economic feasibility under the Korean electricity market, which is cost-based trading system, is studied to find out institutional support system. The results of feasibility summarized that under the current electricity trading system, if the IGCC is considered like a conventional plant such as nuclear or coal-fired power plants, it will not be expected that its investment will be recouped within life-time. The reason is that the availability of an IGCC will plummet since 2016 when several nuclear and coal-fired power plants will be constructed additionally. To ensure the reasonable return on investment (NPV>0 IRR>Discount rate), the availability of IGCC should be higher than 77%. To do so, the current electricity trading system is amended that the IGCC generator must be considered as renewable generators to set up Price Setting Schedule and it should be considered as pick load generators, not Genco's coal fired-generators, in the Settlement Payment.