• 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.

• 한국재난정보학회 논문집
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• 제18권2호
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• pp.333-342
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• 2022

연구목적: 중소기업에서 「안전보건경영시스템」의 유지와 실행만으로 「중대재해처벌법」의 요구사항에 적합하도록 「중대재해처벌법」과 「안전보건경영시스템」연계방안을 제시하고자 한다. 연구방법: 문헌조사와 같이 이론적 접근을 수행한다. 먼저 이론적 접근으로 「중대재해처벌법」의 안전보건관리체계를 살펴보고, 「안전보건경영시스템」의 요구사항을 분석하였으며, 그리고 2024년 「중대재해처벌법」이 적용되는 5인 이상 49인까지의 중소기업의 통계와 사고 사망자수를 조사하였다. 연구결과: 「안전보건경영시스템 (ISO 45001)」요구사항과 「중대재해처벌법」의 안전보건관리체계의 연계하는데 문제점이 발견되지는 않았다. 결론: 「안전보건경영시스템」인증을 받기 위한 시스템 구축과 실행으로 「중대재해처벌법」의 안전보건관리체계의 요구사항을 충족할 수 있으리라고 판단된다.

• 한국건설안전학회 논문집
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• 제5권1호
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• pp.25-30
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• 2023

2022년 1월 27일부터 시행된 중대재해처벌법은 안전조치를 소홀히 한 사업주나 관리자를 중대재해 발생 시 1년 이상의 징역형에 처하도록 하고 있다. 이 중대재해의 처벌에 관한 법률이 올해부터 시행되면서 사회의 많은 관심을 받고 있다. 전문가들에 따르면 안전사고율을 낮추는 법안의 당초 취지와 달리 기업에 대한 처벌을 높이는 데 초점을 맞춰 법안을 만들었다는 의견이 대부분이다. 본 연구는 사례연구를 통해 이러한 내용이 사실인지를 확인하였으며, 이를 지속할 경우 기업과 근로자 간의 갈등이 심화되어 안전사고율이 감소하지 않을 것이다. 따라서 본 연구는 기업의 처벌을 무조건 늘리는 것이 아니라 설문조사를 통해 많은 상황을 비교하고 기업과 근로자가 안전사고율을 낮추기 위해 중대재해처벌법에 협력할 수 있는 방안을 제시하였다. 본 연구는 안전한 사회 건설을 위한 발전된 중대재해 처벌 법의 제정을 위한 기초자료로 활용될 것으로 사료된다.

• 한국철도학회:학술대회논문집
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• 한국철도학회 2003년도 추계학술대회 논문집(III)
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• pp.552-557
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• 2003

The catenary system used in urban electrification railway have suffered mechanical and chemical stresses. These brought out severe messenger wire problems, these have caused accident. Especially, messenger wire is severely influenced by corrosion, vibration and tension. This paper investigate for the cause analysis of broken accident on messenger wire in catenary system. To analyze the cause analysis of accident, we have conducted experiment such as XRD, SEM, EDX and tensile test.

• Nuclear Engineering and Technology
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• 제44권3호
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• pp.225-236
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• 2012

The Fukushima-Daiichi accident shook the world, as a well-known plant design, the General Electric BWR Mark I, was heavily damaged in the tsunami, which followed the Great Japanese Earthquake of 11 March 2011. Plant safety functions were lost and, as both AC and DC failed, manoeuvrability of the plants at the site virtually came to a full stop. The traditional system of Emergency Operating Procedures (EOPs) and Severe Accident Management Guidelines (SAMG) failed to protect core and containment, and severe core damage resulted, followed by devastating hydrogen explosions and, finally, considerable radioactive releases. The root cause may not only have been that the design against tsunamis was incorrect, but that the defence against accidents in most power plants is based on traditional assumptions, such as Large Break LOCA as the limiting event, whereas there is no engineered design against severe accidents in most plants. Accidents beyond the licensed design basis have hardly been considered in the various designs, and if they were included, they often were not classified for their safety role, as most system safety classifications considered only design basis accidents. It is, hence, time to again consider the Design Basis Accident, and ask ourselves whether the time has not come to consider engineered safety functions to mitigate core damage accidents. Associated is a proper classification of those systems that do the job. Also associated are safety criteria, which so far are only related to 'public health and safety'; in reality, nuclear accidents cause few casualties, but create immense economical and societal effects-for which there are no criteria to be met. Severe accidents create an environment far surpassing the imagination of those who developed EOPs and SAMG, most of which was developed after Three Mile Island - an accident where all was still in place, except the insight in the event was lost. It requires fundamental changes in our present safety approach and safety thinking and, hence, also in our EOPs and SAMG, in order to prevent future 'Fukushimas'.

• 방사성폐기물학회지
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• 제9권4호
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• pp.231-236
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• 2011

지난 2011년 3월의 후쿠시마 원전 사고시 원자로 건물에서의 연쇄적인 수소폭발이 발생하였을 때 관계자들은 제1원전 4호기의 폭발에 더욱 놀랐었는데 이는 그 당시 4호기는 정기보수를 위하여 원자로내 모든 핵연료를 저장조에 보관중이었기 때문이다. 저장조내 냉각수 유실로 노심에서 옮겨진 핵연료가 공기 중에 노출되어 수소가 발생하고 임계가 도달하였다면 더욱 심각할 수도 있기 때문이었는데 다행히 추후에 양호한 냉각수 상태가 확인되어 우려할 상황을 피할 수 있었다. 본 논문에서는 후쿠시마 원전 사고를 계기로 국내 원자력 발전소내 핵연료 임시 저장시설의 안전성과 관련하여 중대사고 관점에서 검토해 보고자 한다.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.3017-3029
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• 2023

In the late in-vessel phase of a nuclear reactor severe accident, the internal heat transfer and crust evolution during the debris bed melting process have important effects on the thermal load distribution along the vessel wall, and further affect the reactor pressure vessel (RPV) failure mode and the state of melt during leakage. This study coupled the phase change model and large eddy simulation to investigate the variations of the temperature, melt liquid fraction, crust and heat flux distributions during the debris bed melting process in the hypothetical severe accident of HPR1000. The results indicated that the heat flow towards the vessel wall and upper surface were similar at the beginning stage of debris melting, but the upward heat flow increased significantly as the development of the molten pool. The maximum heat flux towards the vessel wall reached 0.4 MW/m². The thickness of lower crust decreased as the debris melting. It was much thicker at the bottom region with the azimuthal angle below 20° and decreased rapidly at the azimuthal angle around 20-50°. The maximum and minimum thicknesses were 2 and 90 mm, respectively. By contrast, the distribution of upper crust was uniform and reached stable state much earlier than the lower crust, with the thickness of about 10 mm. Moreover, the sensitivity analysis of initial condition indicated that as the decrease of time interval from reactor scram to debris bed dried-out, the maximum debris temperature and melt fraction became larger, the lower crust thickness became thinner, but the upper crust had no significant change. The sensitivity analysis of in-vessel retention (IVR) strategies indicated that the passive and active external reactor vessel cooling (ERVC) had little effect on the internal heat transfer and crust evolution. In the case not considering the internal reactor vessel cooling (IRVC), the upper crust was not obvious.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4514-4521
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• 2022

In this study, the effects of heat and radiation on the degradation behaviour of fluoroelastomer under simulated normal operation and a severe accident environment were investigated using sequential testing of gamma irradiation and thermal degradation. Tensile properties and Shore A hardness were measured, and thermogravimetric analysis was used to evaluate the degradation behaviour of fluoroelastomer. Fourier transform infrared spectroscopy and X-ray photoelectron spectroscopy were used to characterize the structural changes of the fluoroelastomer. Heat and radiation generated in nuclear power plant break and deform the chemical bonds, and fluoroelastomer exposed to these environments have decreased C-H and functional groups that contain oxygen and double bonds such as C-O, C=O and C=C were generated. These functional groups were formed by auto oxidation by reacting free radicals generated from the cleaved bond with oxygen in the atmosphere. In this auto oxidation reaction, crosslinks were generated where bonded to each other, and the mobility of molecules was decreased, and as a result, the fluoroelastomer was hardened. This hardening behaviour occurred more significantly in the severe accident environment than in the normal operation condition, and it was found that thermal stability decreased with the generation of unstable structures by crosslinking.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.162-176
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• 2022

In Lead-based reactor (LBR) severe accident, the meltdown and migration inside the reactor core will lead to fuel fragment concentration, which may further cause re-criticality and even core disintegration. Accurately predicting the migration and solidification behavior of melt in LBR severe accidents is of prime importance for safety analysis of LBR. In this study, the Moving Particle Semi-implicit (MPS) method is validated and used to simulate the migration and solidification behavior. Two main surface tension models are validated and compared. Meanwhile, the MPS method is validated by the L-plate solidification test. Based on the improved MPS method, the migration and solidification behavior of melt in LBR severe accident was studied furthermore. In the Pb-Bi coolant, the melt flows upward due to density difference. The migration and solidification behavior are greatly affected by the surface tension and viscous resistance varying with enthalpy. The whole movement process can be divided into three stages depending on the change in velocity. The heat transfer of core melt is determined jointly by two heat transfer modes: flow heat transfer and solid conductivity. Generally, the research results indicate that the MPS method has unique advantage in studying the migration and solidification behavior in LBR severe accident.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1174-1183
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• 2016

Hydrogen release during severe accidents poses a serious threat to containment integrity. Mitigating procedures are necessary to prevent global or local explosions, especially in large steel shell containments. The management of hydrogen safety and prevention of over-pressurization could be implemented through a hydrogen reduction system and spray system. During the course of the hypothetical large break loss-of-coolant accident in a nuclear power plant, hydrogen is generated by a reaction between steam and the fuel-cladding inside the reactor pressure vessel and also core concrete interaction after ejection of melt into the cavity. The MELCOR 1.8.6 was used to assess core degradation and containment behavior during the large break loss-of-coolant accident without the actuation of the safety injection system except for accumulators in Beznau nuclear power plant. Also, hydrogen distribution in containment and performance of hydrogen reduction system were investigated.