• Journal of Radiation Protection and Research
• /
• v.6 no.1
• /
• pp.45-52
• /
• 1981

The March 1979 accident at Three Mile Island provided physicians specializing in radiation medicine an opportunity to observe the field under conditions never seen before. Since no, injuries occurred at the site or within the community, medical personnel were immediately involved in efforts to allay fear, provide accurate information, and replace labortory resources rendered ineffective by the release in the reactor building. Valuable insights concerning medical emergency planning are derived from the accident; suggestions are made for handling any future mishaps.

• Journal of the KSME
• /
• v.20 no.6
• /
• pp.448-456
• /
• 1980

현재와 같이 생산, 저장, 수송, 에너지 수요등 모든 측면에서 대용량화되고, 여기서 부수되는 기기 및 구조물이 대형화된 상태에서의 사고가 어느정도의 심핵성을 지니며, 어떠한 규모로 손실을 초래하는가 하는 것은, 지금까지의 여러사고예를 수록하고 있는 외국문헌을 참조하지 않더라도, 작년 국내 신문지상에서도 보도된 바 있는 미국의 Three Mile Island의 원자력발전소의 사고 (1979, 4)의 내용을 상기하면 거의 추측이 가능하리라 생각이 된다.

• 방재와보험
• /
• s.30
• /
• pp.64-67
• /
• 1986

우리나라 뿐만 아니라 전세계를 경악케 하였던 소련의 체르노빌 핵발전소 방사능 누출사고를 계기로 최근 우리들의 관심사가 되고있는 원자력발전소 사고사례를 게재한다. 이 내용은 미국 안전공학사협회 (American Society of Safety Engineers)가 발행하는 "Professional Safety"지 1979년 8월호에 실린 Three Mile Island원자력발전소의 사고내용을 번역한 것이다.

• Nuclear Engineering and Technology
• /
• v.45 no.5
• /
• pp.625-636
• /
• 2013

Pressing demands of economic competitiveness, the need for large-scale deployment, minimizing the need of human intervention, and experience from the past events and incidents at operating reactors have guided the evolution and innovations in reactor technologies. Indian innovative reactor 'AHWR' is a pressure-tube type natural circulation based boiling water reactor that is designed to meet such requirements, which essentially reflect the needs of next generation reactors. The reactor employs various passive features to prevent and mitigate accidental conditions, like a slightly negative void reactivity coefficient, passive poison injection to scram the reactor in event of failure of the wired shutdown systems, a large elevated pool of water as a heat sink inside the containment, passive decay heat removal based on natural circulation and passive valves, passive ECC injection, etc. It is designed to meet the fundamental safety requirements of safe shutdown, safe decay heat removal and confinement of activity with no impact in public domain, and hence, no need for emergency planning under all conceivable scenarios. This paper examines the role of the various passive safety systems in prevention and mitigation of severe plant conditions that may arise in event of multiple failures. For the purpose of demonstration of the effectiveness of its passive features, postulated scenarios on the lines of three major severe accidents in the history of nuclear power reactors are considered, namely; the Three Mile Island (TMI), Chernobyl and Fukushima accidents. Severe plant conditions along the lines of these scenarios are postulated to the extent conceivable in the reactor under consideration and analyzed using best estimate system thermal-hydraulics code RELAP5/Mod3.2. It is found that the various passive systems incorporated enable the reactor to tolerate the postulated accident conditions without causing severe plant conditions and core degradation.

• Journal of the Korean Institute of Gas
• /
• v.10 no.2 s.31
• /
• pp.47-54
• /
• 2006

The immediate causes of accidents are often identified as human error or technical failure but the investigation and analysis of the circumstances surrounding major industrial accidents such as Three Mile Island, Chemobyl and Kings Cross have revealed issues beyond the immediate causes. These issues relate to wider considerations of the safety culture. The safety culture of an organization is very complex and hard to study, but it is possible to examine norms that make up the culture. This paper focuses on environmental attitudes and actions among managerial and non-managerial workers in high risk industry such as chemical industries. The main purpose is to get a better understanding of safety culture and to develop measuring tool by examining their nature and strength and by analysing underlying factors that offer explanations for attitudinal differences.

• International Journal of Safety
• /
• v.11 no.1
• /
• pp.1-5
• /
• 2012

It has been reported that the plugged steam generator tube of Three Mile Island Unit 1 in America was damaged by growing flaw and then this steam generator tube destroyed the nearby steam generator tubes of normal state. On this account, stabilizer installation is necessary to prevent secondary damage of the steam generator tubes. The flow-induced vibration is one of the major causes of the fluid-elastic instability. To guarantee the structural integrity of steam generator tubes, the flow-induced vibration caused by the fluid-elastic instability is necessary to be suppressed. In this paper, the effective velocity and the critical velocity are calculated to evaluate the fluid-elastic instability. In addition, stability ratio value of the steam generator tubes is evaluated in order to propose one criterion when to determine stabilizer installation.

• Proceedings of the Korean Nuclear Society Conference
• /
• 1998.05a
• /
• pp.818-823
• /
• 1998

Three Mile Island Unit-2 (TMI-2)의 사고 후 OECD-NEA 주관의 연구에 의하면 압력용기 하부의 노즐이 국부열점(hot spot) 영역의 경우 거의 압력용기 바닥까지 용융되었음이 조사되었다. [1]. 이러한 재배치된 용융노심의 열속에 의하여 압력용기의 외부와 통하는 penetration tube weld(노즐 용접부)가 파손된다면 내부의 고압상태로 인해 penetration tube ejection 사고 및 이에 따르는 용융노심의 압력용기 외부로의 유출 가능성까지 배제할 수 없을 것이다. 본 연구의 출발점은 중대사고시 이러한 압력 및 열속에 따르는 노즐 용접부의 파손확률을 결정하는데 있다. 크리프 파출시 기존의 해석에서 쓰인 deterministic approach를 개선하여 probabilistic approach를 개발하였다. 또한 기존의 해석에서 쓰인 단순한 안전 여유도(margin-to-failure)의 개념과 비교하여 용접부에서의 파손확률을 계산하였다.

• Journal of the Korean Society of Safety
• /
• v.29 no.6
• /
• pp.172-178
• /
• 2014

In terms of an accident management, the cases causing severe core damage need to be analyzed and arranged systematically for an easy access to the results since the Three Mile Island (TMI) accident. The objectives of this paper are to explain how to identify the plant response and cope with its vulnerabilities using the probabilistic safety assessment (PSA) quantified results and severe accident database SARDB(Severe Accident Risk Data Bank) based on sequences analysis results. Although PSA has been performed for the Korean Standard Power Plants (KSNPs), and that it considered the necessary sequences for an assessment of the containment integrity. The developed Database (DB) system includes a graphical display for a plant and equipment status, previous research results by a knowledge-based technique, and the expected plant behaviour. The plant model used in this paper is oriented to the cases of loss of coolant accident (LOCA) is be used as a training simulator for a severe accident management.

• Nuclear Engineering and Technology
• /
• v.44 no.4
• /
• pp.379-386
• /
• 2012

A symptom-based procedure with a critical safety function monitoring system has been established to reduce the operator's diagnosis and cognitive burden since the Three-Mile Island (TMI) accident. However, it has been reported that a symptom-based procedure also requires an operator's cognitive efforts to cope with off-normal events. This can be caused by mismatches between a static model, an emergency operating procedure (EOP), and a dynamic process, the nature of an ongoing situation. The purpose of this study is to share the evidence of mismatches that may result in an excessive cognitive burden in conducting EOPs. For this purpose, we analyzed simulated emergency operation records and observed some operator behaviors during the EOP operation: continuous steps, improper description, parameter check at a fixed time, decision by information previously obtained, execution complexity, operation by the operator's knowledge, notes and cautions, and a foldout page. Since observations in this study are comparable to the results of an existing study, it is expected that the operational behaviors observed in this study are generic features of operators who have to cope with a dynamic situation using a static procedure.

• Nuclear Engineering and Technology
• /
• v.54 no.11
• /
• pp.4272-4279
• /
• 2022

In this study, a cross section stochastic sampling (S.S.) capability is implemented into both the McCARD continuous energy Monte Carlo code and MIG multiple-correlated data sampling code. The ENDF/B-VII.1 covariance data based 30 group cross section sets and the SCALE6 covariance data based 44 group cross section sets are sampled by the MIG code. Through various uncertainty quantification (UQ) benchmark calculations, the McCARD/MIG results are verified to be consistent with the McCARD stand-alone sensitivity/uncertainty (S/U) results and the XSUSA S.S. results. UQ analyses for Three Mile Island Unit 1, Peach Bottom Unit 2, and Kozloduy-6 fuel pin problems are conducted to provide the uncertainties of k_eff and microscopic and macroscopic cross sections by the McCARD/MIG code system. Moreover, the SNU S/U formulations for uncertainty propagation in a MC depletion analysis are validated through a comparison with the McCARD/MIG S.S. results for the UAM Exercise I-1b burnup benchmark. It is therefore concluded that the SNU formulation based on the S/U method has the capability to accurately estimate the uncertainty propagation in a MC depletion analysis.