• Proceedings of the Safety Management and Science Conference
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• 2012.04a
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• pp.447-460
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• 2012

현재 위험물 운송 사고는 점차 감소하는 추세이지만 위험물 운송 사고는 특성상 한번의 사고로 인해 주변의 인명 및 환경에 막대한 피해를 입히게 된다. 이러한 위험물 운송 사고의 대응을 위해 각 기관들이 있지만 기관별로 메뉴얼 및 방침이 다르게 적용되어 있어 운송 사고에 대한 정확한 정보전달이 미비한 실정이다. 결과적으로 운송사고에 대한 정보가 부족하여 잘못된 대응계획으로 사고처리가 늦어질 수도 있다. 사고시 적절한 대응을 위해서는 위험물질의 종류와 양, 사고위치, 주변상황 등에 대한 정보가 전달되어야 사고대응에 대한 장비, 인력 편성 등 정확한 대응이 이루어질 수 있다. 현재 국내의 정부 유관기관별로 위험물질 사고대응메뉴얼을 구축하여 제공하고 있으나, 운송 및 수송 사고에 대한 사고대응메뉴얼의 부재로 2차적인 피해가 발생되고 있다. 따라서 본 논문에서는 위험물질의 표준화하고 이에 대한 사고대응매뉴얼을 개발하여 위험물 운송 사고시 사고에 대한 정보를 유관기관에게 신속 정확하게 전파하여 통합적인 초동대응이 이루어질 수 있도록 하였다.

• Journal of the Korean Society of Safety
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• v.27 no.3
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• pp.28-35
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• 2012

We investigated the breakdown patterns of silica glass under the various impacts by forensic scientific analysis. The impacts were chosen by thermal and projectiles impact stress. Thermal impacts under the fire were constructed by changing the position of the flame with gas torches. Physical impacts were constructed with the projectiles of slingshots and rifles by changing the size of the projectiles at the surfaces and the initial distance. Also we identified incident angles of the projectiles by analysis of the punching portion at the glass surface. Under the thermal impacts, the breakdown patterns were various by changing the position of the flame. Especially, the configuration of breakdown patterns had radial shape with the position of the flame at the center of the glass sheet. Under the physical impacts by the projectiles, the breakdown patterns had always radial shape.

• Journal of the Korea Safety Management & Science
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• v.16 no.2
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• pp.19-29
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• 2014

Korean societal concern for the train accidents is fast and widely increasing with an ever-increasing demand and use for KTX. Most of these train accidents are inclined to be caused by human error. Experts used to attribute the causes of human error to the defects in various aspects such as technology, organizational system, practices, corporate culture, and/or human resource itself. Among the diverse causes of human error, an important one, even though it was rarely focused, may be the issue of impact of rule or procedure change on human error. Giving attention to the implicit importance of this issue, this study intends to highlight the theme of frequent procedure change in railway driving manual as a critical factor of human error. To attain this purpose mentioned above, dual methodologies were adopted. One is to qualitatively analyze the real cases of procedure change in relevant manuals followed by the incident case(passing the station scheduled to stop) happened lately. Another is to quantitatively perform statistical analysis based on questionnaires received from 224 train drivers. Results show that frequent changes in internal affairs procedure is or may be an important factor causing stress and human error from train drivers.

• International Journal of Safety
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• v.10 no.2
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• pp.21-26
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• 2011

In 2001, an independent official board was constituted in Japan to investigate aircraft and railway accidents. In the past 10 years, many accidents and serious incidents have been investigated and these official reports were published by the board, on which the author had sat for 9 years as boarding member. In the interim, there were several train disasters which mocked our trust in railways and also many apparent trivial incidents. In recent years, serious incidents, which a door of running rail cars opens suddenly with some trouble, happen 2 or 3 times in a year. For the past 10 years, such incidents have happened 14 times and 13 cases of them were closed by the board mentioned above. In these 13 cases, no one fell off the rail car, so that the death toll was none luckily. In this paper, these 13 serious incidents are picked up among all the reports published by the board and outlined using some tables. Especially, fall accidents of passengers are discussed mainly from the view point of impact force and duration time. Then, the equation of HIC (Head Injury Criteria) and the risk curves in terms of the HIC are dealt with properly.

• Journal of the Korean Society of Safety
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• v.22 no.6
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• pp.46-54
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• 2007

The purpose of this study was to examines the burning rate of fire retardant treated wood in the cone heater with a one-dimensional integral model. The wood samples used in this study were four species. The species of woods are Redwood, White oak, Douglas fir and Maple. Each sample was nominally 50mm thick and 100mm square. Samples were exposed to a range of incident heat fluxes 10 to $35kW/m^2$ using the cone heater. A one-dimension integral model has been used to predict burning rate, heat of gasification, flame heat fluxes, charring rate and char depth of samples. As a result measurement of mass loss rate, softwoods(Redwood and Douglas fir) has relatively low value than those for hardwoods(White oak and Maple). Average charring rate of woods in case of fire retardant treatment showed reduction effect of 41.29%, 50.00%, 48.18% and 60.82% for Redwood, Douglas fir, White fir and Maple, respectively. Almost all the predictions from integral model showed faster charring than those measured. Average difference between predictions and experimental data was 16%, 9.5% and 11.8% for N, F1 and F2 respectively. Water-soluble fire retardant used in this study find out more effect in hardwood than softwood from the result of measurement of mass loss rate and average charring rate.

• Journal of the Korean Society of Safety
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• v.23 no.6
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• pp.84-90
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• 2008

The mass loss rate and heat release rate of puzzle mats were analysed using variable external irradiation level. Five samples of puzzle mat were tested in this study : Type A, B, C, D and E. Type A, B and C are all general grades whereas Type D and E are both Flame retardant grades. Incident heat fluxs of $25kW/m^2$, $35kW/m^2$, $50kW/m^2$ and $70kW/m^2$ were selected for these experiments. All samples were tested in the horizontal orientation and were wrapped in a single layer of aluminum foil. Each sample was nominally 20mm thick and 100mm square. The combustion heat and mass loss rate were carried out from Oxygen bomb calorimeter and mass loss calorimeter according to ISO 5660-1 respectively. Heat release rates were calculated using the equation ${\dot{Q}}=A_f{\dot{m}}"_X{\Delta}H_c=0.75A_f{\dot{m}}"{\Delta}H_c$. where $A_f$ is the horizontal burning area of the sample, $\dot{m}"$ is mass loss rate per unit area, ${\Delta}H_c$ is complete heat of combustion and 0.75 is combustion efficiency.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.429-436
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• 2022

Investigations of retrospective dosimetry have shown that components of mobile phones are suitable as emergency dosimeters in case of radiological incidents. For physical dosimetry, components can be read out using optically stimulated luminescence (OSL), thermoluminescence (TL) and phototransferred thermoluminescence (PTTL) methods to determine the absorbed dose. This paper deals with a feasibility study of display glass from modern mobile phones that are measured by thermally assisted (Ta) optically stimulated luminescence. Violet (VSL, 405 nm) and infrared (IRSL, 850 nm) LEDs were used for optical stimulation and two protocols (Ta-VSL and Ta-IRSL) were tested. The aim was to systematically investigate the luminescence properties, compare the results to blue stimulated Ta-BSL protocol (458 nm) and to develop a robust measurement protocol for the usage as an emergency dosimeter after an incident with ionizing radiation. First, the native signals were measured to calculate the zero dose signal. Next, the reproducibility and dose response of the luminescence signals were analyzed. Finally, the signal stability was tested after the storage of irradiated samples at room temperature. In general, the developed Ta-IRSL and Ta-VSL protocols indicate usability, however, further research is needed to test the potential of a new protocol for physical retrospective dosimetry.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2445-2453
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• 2021

Since the nuclear data forms a vital component in reactor physics computations, the nuclear community needs processing codes as tools for translating the Evaluated Nuclear Data Files (ENDF) to simulate nuclear-related problems such as an ACE format that is used for MCNP. Errors, inaccuracies or discrepancies in library processing may lead to a calculation that disagrees with the experimentally measured benchmark. This paper provides an overview of the processing and preparation of ENDF/B-VIII.0 incident neutron data with NECP-Atlas and NJOY codes for implementation in the MCNP code. The resulting libraries are statistically inter-compared and tested by conducting benchmark calculations, as the mutualcomparison is a source of strong feedback for further improvements in processing procedures. The database of the benchmark experiments is based on a selection taken from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP handbook) and those proposed by Russell D. Mosteller. In general, there is quite good agreement between the NECP-Atlas1.2 and NJOY21^(1.0.0.json) results with no substantial differences, if the correct input parameters are used.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1127-1133
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• 2021

A Spent Fuel Pool (SFP) is designed to store spent fuel assemblies in the pool. And, a SFP cooling and cleanup system cools the SFP coolant through a heat exchanger which exchanges heat with component cooling water. If the cooling system fails or interfacing pipe (e.g., suction or discharge pipe) breaks, the cooling function may be lost, probably leading to fuel damage. In order to prevent such an incident, it is required to properly cool the spent fuel assemblies in the SFP by either recovering the cooling system or injecting water into the SFP. Probabilistic safety assessment (PSA) is a good tool to assess the SFP risk when an initiating event for the SFP occurs. Since PSA has been focused on reactor-side so far, it is required to study on the framework of PSA approach for SFP and identify the key factors in terms of fuel damage frequency (FDF) through a case study. In this study, therefore, a case study of SFP-PSA on the basis of design information of APR-1400 has been conducted quantitatively, and several sensitivity analyses have been conducted to understand the impact of the key factors on FDF.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3229-3240
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• 2023

Due to neutron radiation, atomic displacement has a significant effect on material in nuclear reactors. A range of secondary displacement models, including the Kinchin-Pease (K-P), Lindhard, Norgett-Robinson-Torrens (NRT), and athermal recombination-corrected displacement per atom (arc-dpa) have been suggested to calculate the number of displacement per atom (dpa). As neutron elastic interaction is the main cause of displacement damage, the focus of the current study is to calculate the atomic displacement caused by the neutron elastic interaction in order to estimate the exact amount of yield strength in a WWER-1000 reactor pressure vessel. To achieve this purpose, the reactor core is simulated by MCNPX code. In addition, a program is developed to calculate the elastic radiation damage induced by the incident neutron flux (RADIX) based on different models using Fortran programming language. Also, due to non-elastic interaction, the displacement damage is calculated by the HEATR module of the NJOY code. ASME E-693-01 standard, SPECTER, NJOY codes, and other pervious findings have been used to validate RADIX results. The results showed that the RADIX(arc-dpa)/HEATR outputs have appropriate accuracy. The relative error of the calculated dpa resulting from RADIX(arc-dpa)/HEATR is about 8% and 46% less than NJOY code, respectively in the ¼ and ¾ vessel wall.