• 대한토목학회논문집
• /
• 제30권6D호
• /
• pp.587-597
• /
• 2010

교통사고가 발생하면 사고 당사자들은 민 형사적인 문제에서 자유롭지 못하기 때문에, 교통사고 조사자는 사고 상황을 정확하게 재현 또는 분석을 하여야 한다. 또한, 이러한 교통사고 발생과 관련한 요인에 대한 분석을 통해 얻어진 자료를 활용하여 교통사고 다발지역의 개선 및 보완작업을 시행하게 된다. 현재까지 알려진 바로는 수많은 교통사고, 교통시설물, 도로설계 등과 관련하여 가장 많은 영향을 미치는 요인은 차량의 속도와 가속능력, 제동능력 등이다. 이는 자동차의 성능과 노면의 마찰계수가 가장 밀접한 영향을 미치는 부분이다. 특히, 사고 순간의 속도의 추정은 교통사고처리특례법의 11개 주요항목인 과속과 관련하여 매우 중요한 사항이기에 정확성이 요구되는 부분이다. 하지만, 국내에서는 아직 이러한 부분에 대한 심도 있는 연구가 많이 진행되지 못하는 것이 현실이다. 이러한 현실을 반영하여 본 연구에서는 차량의 급제동에서 제동흔적이 발생되기 시작할 때까지의 시간인 제동경과시간을 정밀가속도계(Vericom VC2000PC)로 측정하여 제동경과시간과 노면의 마찰계수를 정확히 추정하였다. 실험결과를 분석하여, 여러 가지 특수 아스팔트 포장 및 미끄럼방지포장 종류에 따른 제동경과시간과 마찰계수를 계산하여 연구의 목적에 맞도록 기초자료를 제공하고자 하였다.

• 한국안전학회지
• /
• 제31권6호
• /
• pp.129-134
• /
• 2016

Standard design of APR+(advanced power reactor plus) was certified at 2014 by Korea regulatory body. Based on the experience gained from OPR1000 and APR1400, the APR1400 was being developed as a 1,500MWe class reactor using Korean technologies for design code, reactor coolant pump, and man-machine interface system. APR+ has been basically designed to have the seismic design basis of safe shutdown earthquake (SSE) 0.3g, a 4-train safety concept based on N+2 design philosophy, and a passive auxiliary feedwater system (PAFS). Also, safety issues on the Fukushima-type accidents have been extensively reviewed and applied to enhance APR+ safety. APR+ provides higher reliability and safety against tsunami and earthquake. The purpose of this paper is to implement probabilistic safety assessment considering these design features and to analyze sensitivity of core damage frequency for large loss of coolant accident of APR+.

• 기술사
• /
• 제44권6호
• /
• pp.40-44
• /
• 2011

Nuclear energy in Korea began in 1958, when the Korea's atomic energy act was formulated and the relevant organizations were founded. Since then, notwithstanding the two catastrophe like TMI and Chernobyl accident, Korea made a wise decision to expand the peaceful uses of the nuclear energy as well as to localize the essential nuclear design technology of fuel and nuclear steam supply system. This decision resulted in the success of export of nuclear power plants as well as research reactor in 2010s. The Korea's nuclear policy, which well utilized 'international crisis in nuclear business' as 'opportunity of Korea to get. nuclear technology', is believed nice policy as a role model of nuclear new-comer countries. Based upon the success story of localization of nuclear technology, Korea had an eye for a niche market, which was a basis of development of SMART, Korea-made integral PWR. The operation of a SMART plant can sufficiently provide not only electricity but also fresh water for 100,000 residents. Last two years, Korea's nuclear industry team led by the Korea Atomic Energy Research Institute completed the standard design of SMART and applied to the Korea's regulatory body for standard design approval. Now the Korea's licensing authority is reviewing the design with the relevant documents, and the design team is doing its best to realize its hope to get the approval by the end of this year. From next year, the SMART business including construction and export will be explored by the KEPCO consortium.

• Nuclear Engineering and Technology
• /
• 제43권5호
• /
• pp.413-420
• /
• 2011

In Korea, although the concept of dry storage system for PWR spent fuels first emerged in the early 1990s, wet storage inside nuclear reactor buildings remains the dominant storage paradigm. Furthermore, as the amount of discharged fuel from nuclear power plants increases, nuclear power plants are confronted with the problem of meeting storage capacity demand. Various measures have been taken to resolve this problem. Dry storage systems along with transportation of spent fuel either on-site or off-site are regarded as the most feasible measure. In order to develop dry storage and transportation system safety analyses, development of design techniques, full scale performance tests, and research on key material degradation should be conducted. This paper deals with two topics, structural analysis methodology to assess cumulative damage to transportation packages and the effects of an aircraft engine crash on a dual purpose cask. These newly emerging issues are selected from among the many technical issues related to the development of transportation and storage systems of spent fuels. In the design process, appropriate analytical methods, procedures, and tools are used in conjunction with a suitably selected test procedure and assumptions such as jet engine simulation for postulated design events and a beyond design basis accident.

• 한국안전학회지
• /
• 제29권6호
• /
• pp.166-171
• /
• 2014

Nuclear power plants are equipped with the reactor trip system (RTS) and the engineered safety features actuation system (ESFAS) to improve safety on the normal operation. In the event of the design basis accident (DBA), a various of post accident monitor(PAM)systems support to provide important details (e.g. Containment pressure, temperature and pressure of reactor cooling system and core exit temperature) to determine action of main control room (MCR). Operator should be immediately activated for the accident mitigation with the information. Especially, core exit temperature is a critical parameter because the operating mode converts from normal mode to emergency mode when the temperature of core exit reaches $649^{\circ}C$. In this study, uncertainty which was caused by exterior environment, characteristic of thermocouple/connector and accuracy of calibrator/indicator was evaluated in accordance with ANSI-ISA 67.04. The square root of the sum of square (SRSS) methodology for combining uncertainty terms that are random and independent was used in the synthesis. Every uncertainty that may exist in the hardware which is used to measure the core exit temperature was conservatively applied and the associative relation between the elements of uncertainty was considered simultaneously. As a result of uncertainty evaluation, the channel statistical allowance (CSA) of single channel of core exit temperature was +1.042%Span. The range of uncertainty, -0.35%Span ($-4.05^{\circ}C$) ~ +2.08%Span($24.25^{\circ}C$), was obtained as the operating criteria of core exit temperature.

• 토지주택연구
• /
• 제7권3호
• /
• pp.147-156
• /
• 2016

정부는 서민들의 주거안정을 위해 1989년부터 공공임대 주택을 지속적으로 공급하고 있다. 하지만 노후 공공임대 주택들은 시설물의 노후화로 인해 안전사고가 발생할 가능성이 높은 편이다. 따라서 본 연구에서는 공공 임대주택단지 내에서 발생하는 생활안전 사고의 유형을 조사하고, 각 유형별 위해요인을 분석하였다. 본 연구의 진행방법은 먼저 선행연구와 인터뷰조사를 통해 아파트단지 내에서 발생하는 생활안전 사고의 유형을 파악하였다. 그리고 설문조사를 통해 공공 임대주택단지에서 주로 발생하는 생활안전 사고의 유형을 파악하고, 사고유형별 생활안전 위해요인을 분석하였다. 본 연구의 결과는 공공임대주택 거주자들이 안전하고 편안한 주거생활을 위해 계획적이고 체계적인 관리를 위한 계획관리의 기본정보가 될 것이며, 향후 신규 임대주택공급 시 설계 및 관리방법 개선의 기초정보로 유용할 것이다.

• Nuclear Engineering and Technology
• /
• 제40권3호
• /
• pp.199-212
• /
• 2008

A small-break loss of coolant accident (SB-LOCA) test with a break size equivalent to a 3-inch cold leg break of the APR1400 was carried out as the first transient integral effect test using the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation). This was the first integral effect test to investigate the integral performance of the test facility and to verify its simulation capability for one of the design-basis accidents. Reasonably good thermal hydraulic data was obtained so that an integral performance of the fluid sub-systems was identified and control performance of the ATLAS was confirmed under real thermal hydraulic conditions. Based on the measured data, a post-test calculation was carried out using the best-estimate thermal hydraulic safety analysis code, MARS 3.1, and the similarity between the expected and actual data was investigated. On the whole, the post-test calculation reasonably predicts the major thermal hydraulic parameters measured during the SB-LOCA test. The obtained data will be used to enhance the simulation capability of the ATLAS and to improve an input model of the ATLAS for simulation of other target scenarios.

• Nuclear Engineering and Technology
• /
• 제45권7호
• /
• pp.871-894
• /
• 2013

KAERI (Korea Atomic Energy Research Institute) has been operating an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), for transient and accident simulations of advanced pressurized water reactors (PWRs). Using ATLAS, a high-quality integral effect test database has been established for major design basis accidents of the APR1400 plant. A Domestic Standard Problem (DSP) exercise using the ATLAS database was promoted to transfer the database to domestic nuclear industries and contribute to improving a safety analysis methodology for PWRs. This $2^{nd}$ ATLAS DSP (DSP-02) exercise aims at an effective utilization of an integral effect database obtained from ATLAS, the establishment of a cooperation framework among the domestic nuclear industry, a better understanding of the thermal hydraulic phenomena, and an investigation into the possible limitation of the existing best-estimate safety analysis codes. A small break loss of coolant accident with a 6-inch break at the cold leg was determined as a target scenario by considering its technical importance and by incorporating interests from participants. This DSP exercise was performed in an open calculation environment where the integral effect test data was open to participants prior to the code calculations. This paper includes major information of the DSP-02 exercise as well as comparison results between the calculations and the experimental data.

• Nuclear Engineering and Technology
• /
• 제55권2호
• /
• pp.696-706
• /
• 2023

Consequences of an anticipated Beyond Design Basis Accident (BDBA) Long-Term Station Blackout (LTSBO) event with complete loss of grid power in the VVER-1200 reactor of Rooppur Nuclear Power Plant (NPP) of Unit-1 are assessed using the RASCAL 4.3 code. This study estimated the released radionuclides, received public radiological dose, and ground surface concentration considering 3 accident scenarios of International Nuclear and Radiological Event Scale (INES) level 7 and two meteorological conditions. Atmospheric transport, dispersion, and deposition processes of released radionuclides are simulated using a straight-line trajectory Gaussian plume model for short distances and a Gaussian puff model for long distances. Total Effective Dose Equivalent (TEDE) to the public within 40 km and radionuclides contribution for three-dose pathways of inhalation, cloudshine, and groundshine owing to airborne releases are evaluated considering with and without passive safety Emergency Core Cooling System (ECCS) in dry (winter) and wet (monsoon) seasons. Source term and their release rates are varied with the functional duration of passive safety ECCS. In three accident scenarios, the TEDE of 10 mSv and above are confined to 8 km and 2 km for the wet and dry seasons, respectively in the downwind direction. The groundshine dose is the most dominating in the wet season while the inhalation dose is in the dry season. Total received doses and surface concentration in the wet season near the plant are higher than those in the dry season due to the deposition effect of rain on the radioactive substances.

• Nuclear Engineering and Technology
• /
• 제47권7호
• /
• pp.915-923
• /
• 2015

Pyroprocess technology has been considered as a fuel cycle option to solve the spent fuel accumulation problems in Korea. The Korea Atomic Energy Research Institute, Daejeon, Korea has been studying pyroprocess technology, and the conceptual design of an engineering-scale pyroprocess facility, called the Reference Engineering-scale Pyroprocess Facility, has been performed on the basis of a 10 ton heavy metal throughput per year. In this paper the concept of Reference Engineering-scale Pyroprocess Facility is introduced along with its safety requirements for the protection of facility workers, collocated workers, the off-site public, and the environment. For the identification of safety structures, systems, and components and/or administrative controls, the following activities were conducted: (1) identifying hazards associated with operations; (2) identifying potential events associated with these hazards; and (3) identifying the potential preventive and/or mitigative controls that reduce the risk associated with these accident events. This study will be used to perform a safety evaluation for accidents involving any of the hazards identified, and to establish safety design policies and propose a more definite safety design.