• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1963-1973
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• 2020

Liquid metal-cooled reactors use various passive safety systems driven by natural circulation. Investigating these safety systems experimentally is more advantageous by using a simulant. Although numerous experimental approaches have been applied to natural circulation-driven passive safety systems using simulants, there has been no clear validation of the similarity law. To validate the similarity law experimentally, SINCRO-V experiment was conducted using Wood's metal and water for simulant of the Wood's metal. A pair of SINCRO-V facilities with length-scale ratio of 14.1:1 for identical Bo' was investigated, which was the main similarity parameter in temperature field simulation. In the experimental range of 0.2-1.0% of decay heat, the temperature distribution characteristics of the small water facility were very similar to that of the large Wood's metal facility. The temperature of the Wood's metal predicted by the water experiment showed good agreement with the actual Wood's metal temperature. Despite some error factors like discordance of Gr' and property change along the temperature, the water experiment predicted the Wood's metal temperature with an error of 27%. The validity of the similarity law was confirmed by the SINCRO-V experiments.

• 시스템엔지니어링학술지
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• 제12권2호
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• pp.91-99
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• 2016

The safety of nuclear power plants has received much attention; this safety largely depends on the continuous availability of electrical energy source during all modes of nuclear power plant operation. A station blackout (SBO) describes the loss of the off-site electric power, the failure of the emergency diesel generators, and the unavailability of the alternate AC (AAC) power. Consequently, all systems that are AC powered such as the safety injection, shutdown cooling, component cooling water, and essential service water systems are unavailable. The aim of this study is to investigate the deficiencies of the existing alternatives for coping with an extended SBO for APR1400 design. The method is analyzing the existing deficiencies and proposing an optimal solution for the NPP design during the extended SBO. This study, established a new passive system, called passive decay heat removal system (PDHRS), using systems engineering approach.

• 한국화재소방학회논문지
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• 제22권2호
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• pp.97-103
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• 2008

최근 다중이용업소는 사회적 요구 및 이용자의 편리함을 위해 날로 복잡, 다양화함에 따라 일반적인 화재 발생시에도 피난에 장애를 주는 여러 요인이 발생하는 등 피난안전성은 점차 저하되고 있다. 피난계단, 복도 등 Passive system만을 통한 충분한 피난안전성의 확보에는 여러 가지 제약요인이 따르기 때문에 피난설비는 화재시 그 기능을 최대한 발휘할 수 있도록 설치되고, 관리되어야 한다. 본 연구에서는 다중이용업소의 공간 및 이용형태 등 방재적 특성에 맞는 피난설비 및 관련 규정의 최적화를 도모하고자 다중이용업소 피난설비의 효율성 및 화재안전의식 등에 대한설문조사 등을 통하여 다중이용업소의 피난설비의 문제점과 개선방안을 도출하고자 한다.

• 한국안전학회지
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• 제16권4호
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• pp.200-207
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• 2001

회전하는 기계에서 전달되는 조화적인 진동력이 수동 및 능동 진동 격리기를 통하여 중간 지지구조물에 어떻게 전달되는 것인가를 연구하였다. 이를 위하여 이론적인 모델은 모든 축에 대하여 수평과 수직방향의 힘과 모멘트를 고려하여 작성되었으며, 실험은 중간 구조물에 전달되는 회전방향 및 직선방향의 진동을 최소화하기 위하여 2단으로 구성된 중간 지지구조물에 부착된 진동 액츄에이터를 사용하였다. 진동원에 의하여 발생된 진동이 에러 센서에서 측정되었으며 제어원과 에러센서사이의 전달함수가 측정되었다. 1-100Hz사이의 주파수 범위에 있어서 기존의 수동격리기와 직렬로 설치된 능동격리기를 통하여 전달된 힘이 실제로 감소되었음을 실험결과를 통하여 확인하였다.

• 조명전기설비학회논문지
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• 제23권6호
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• pp.31-39
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• 2009

본 논문은 기존의 자동문시스템의 단점을 개선하여 보다 지능적인 기능을 추가함으로써 통행자에게 편의를 제공하고, 전격손실을 낮추는 보다 효율적인 초음파센서 기반의 지능형 자동문 시스템을 제안한다. 기존의 자동문시스템에서는 물체 및 인체 감지 기능 센서로서 원적외선 열선 감지센서와 근적외선 반사식 감지센서를 사용해 왔나. 기존의 방식은 상황에 맞지 않는 문닫힘에 의한 전력손실, 빠르게 다가오는 물체에 대한 적응 부족, 그리고 안전에 대한 문제점이 있었다. 제안된 초음파센서를 이용한 지능형자동문 시스템은 기존의 자동문시스템에 비교하여 필요없는 문닫힘 방지에 의한 전력손실 방지와 빠르게 접근하는 물체에 적응적으로 자동문이 빠르게 열릴 수 있게 하며, 안전도를 높여 기존 방식의 자동문보다 운용적, 경제적, 안정 면에서 성능을 개선하였다.

• Nuclear Engineering and Technology
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• 제52권1호
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• pp.27-36
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• 2020

Natural convection and thermal stratification phenomena are found in large water pools that are being used as heat sinks for decay heat removal from the reactor core using passive heat removal systems. In this study, the two-phase (water and air) natural convection and thermal stratification phenomena with conjugate heat transfer in the rectangular enclosure were investigated numerically using ANSYS Fluent 17.2 code. The transient numerical simulations of these phenomena in the full-scale computational domain of the experimental facility were performed. Generation of water vapour bubbles around the heater rod and evaporation phenomena were included in this numerical investigation. The results of numerical simulations are in good agreement with experimental measurements. This shows that the natural convection is formed in region above the heater rod and the water is thermally stratified in the region below the heater rod. The heat from higher region and from the heater rod is transferred to the lower region via conduction. The thermal stratification disappears and the water becomes well mixed, only after the water temperature reaches the saturation temperature and boiling starts. The developed modelling approach and obtained results provide guidelines for numerical investigations of thermal-hydraulic processes in the water pools for passive residual heat removal systems or spent nuclear fuel pools considering the concreate walls of the pool and main room above the pool.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.163-164
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• 2004

• 대한안전경영과학회지
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• 제24권4호
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• pp.71-83
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• 2022

This study is to investigate the degree of field awareness of how each provision of the Serious Disaster Punishment Act will affect accident prevention. As a result of conducting a survey of construction site workers, it was analyzed that construction site workers had a low sense of safety and did not voluntarily engage in safety activities. And it can be seen that they are taking a very passive position that it is efficient to implement safety under the supervision of the supervisor. Workers who need voluntary construction site safety activities are most aware of the situation at the site, but the problem was pointed out as "lack of procedures for listening to opinions and preparing improvement measures" for these problems. Future research is needed to see if this is a problem for individual workers or for those in charge of safety and health who do not correct it even though active opinions have been submitted.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2499-2508
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• 2021

Passive containment cooling systems (PCCSs) have been actively studied to improve the inherent safety of nuclear power plants. Hered, we present two concepts, open-loop PCCS (OL-PCCS) and closed-loop PCCS (CL-PCCS), applicable to the PWR with a concrete-type containment. We analyzed the heat-removal performance and flow instability of these PCCS concepts using the GOTHIC code. In both cases, PCCS performance improved when a passive containment cooling heat exchanger (PCCX) was installed in the lower part of the containment building. The OL-PCCS was found to be superior in terms of heat-removal performance. However, in terms of flow instability, the OL-PCCS was more vulnerable than the CL-PCCS. In particular, the possibility of flow instability was higher when the PCCX was installed in the upper part of the containment. Therefore, the installation location of the OL-PCCS should be restricted to minimize flow instability. Conversely, a CL-PCCS can be installed without any positional restriction by adjusting the initial system pressure within the loop, which eliminates flow instability. These results could be used as base data for the thermo-hydraulic evaluation of PCCS in PWR with a large dry concrete-type containment.

• 한국안전학회지
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• 제35권4호
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• pp.84-91
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• 2020

The Korean nuclear industry developed the SPACE (Safety and Performance Analysis Code for nuclear power plants) code and this code adpots two-phase flows, two-fluid, three-field models which are comprised of gas, continuous liquid and droplet fields and has a capability to simulate three-dimensional model. According to the revised law by the Nuclear Safety and Security Commission (NSSC) in Korea, the multiple failure accidents that must be considered for accident management plan of nuclear power plant was determined based on the lessons learned from the Fukushima accident. Generally, to improve the reliability of the calculation results of a safety analysis code, verification work for separate and integral effect experiments is required. In this reason, the goal of this work is to verify calculation capability of SPACE code for multiple failure accident. For this purpose, it was selected the experiment which was conducted to simulate a Multiple Steam Generator Tube Rupture(MSGTR) accident with Passive Auxiliary Feedwater System(PAFS) operation by Korea Atomic Energy Research Institute (KAERI) and focused that the comparison between the experiment results and code calculation results to verify the performance of the SPACE code. The MSGR accident has a unique feature of the penetration of the barrier between the Reactor Coolant System (RCS) and the secondary system resulting from multiple failure of steam generator U-tubes. The PAFS is one of the advanced safety features with passive cooling system to replace a conventional active auxiliary feedwater system. This system is passively capable of condensing steam generated in steam generator and feeding the condensed water to the steam generator by gravity. As the results of overall system transient response using SPACE code showed similar trends with the experimental results such as the system pressure, mass flow rate, and collapsed water level in component. In conclusion, it could be concluded that the SPACE code has sufficient capability to simulate a MSGTR accident.