• Nuclear Engineering and Technology
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• 제48권2호
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• pp.545-553
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• 2016

The failure probabilities of the reactor pressure vessel (RPV) for low temperature over-pressurization (LTOP) and cool-down transients are calculated in this study. For the cool-down transient, a pressure-temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME) code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition ($RT_{NDT}$). The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.

• Nuclear Engineering and Technology
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• 제39권4호
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• pp.257-272
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• 2007

Supercritical water-cooled reactors (SCWRs) are recognized as a Generation IV reactor concept. The Super LWR is a pressure-vessel type thermal spectrum SCWR with downward-flow water rods and is currently under study at the University of Tokyo. This paper reviews Super LWR safety. The fundamental requirement for the Super LWR, which has a once-through coolant cycle, is the core coolant flow rate rather than the coolant inventory. Key safety characteristics of the Super LWR inhere in the design features and have been identified through a series of safety analyses. Although loss-of-flow is the most important abnormality, fuel rod heat-up is mitigated by the "heat sink" and "water source" effects of the water rods. Response of the reactor power against pressurization events is mild due to a small change in the average coolant density and flow stagnation of the once-through coolant cycle. These mild responses against transients and also reactivity feedbacks provide good inherent safety against anticipated-transient-without-scram (ATWS) events without alternative actions. Initiation of an automatic depressurization system provides effective heat removal from the fuel rods. An "in-vessel accumulator" effect of the reactor vessel top dome enhances the fuel rod cooling. This effect enlarges the safety margin for large LOCA.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3319-3326
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• 2021

As a form of industrial control systems (ICS), nuclear instrumentation and control (I&C) systems have been digitalized increasingly. This has raised in turn cyber security concerns. Cyber security for ICS is important because cyber-attacks against ICS can cause not only equipment damage and loss of production but also personal and public safety hazards unlike in general IT environments. Numerous risk analyses have been carried out to enhance the safety of ICS and recently, many studies related to the cyber security of ICS are being conducted. Many existing risk analyses and cyber security studies have considered safety and cyber security separately. However, both safety and cyber security perspectives should be considered when analyzing risks for complex and critical ICS facilities such as nuclear power plants (NPPs). In this paper, the STPA-SafeSec methodology is selected to consider both safety and security perspectives when performing a risk analysis for NPPs in order to assess impacts on the safety by cyber-attacks against the digital I&C systems. The STPA-SafeSec methodology was applied to a test-bed system that simulates a condensate water (CD) system in an NPP. The process of the application up to the development of mitigation strategies is described in detail.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.498-508
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• 2021

Micro reactors are increasingly being considered for utilization as distributed power sources. Hence, the probabilistic safety assessment (PSA) of a direct supercritical-CO₂-cooled fast reactor, called micro modular reactor (MMR), was performed in this study; this reactor was developed using innovative design concepts. It adopted a modular design and passive safety systems to minimize site constraints. As the MMR is in its conceptual design phase, design weaknesses and valuable safety insights could be identified during PSA. Level 1 internal event PSA was carried out involving literature survey, system characterization, identification of initiating events, transient analyses, development of event trees and fault trees, and quantification. The initiating events and scenarios significantly contributing to core damage frequency (CDF) were determined to identify design weaknesses in MMR. The most significant initiating event category contributing to CDF was the transients with the power conversion system initially available category, owing to its relatively high occurrence frequency. Further, an importance analysis revealed that the safety of MMR can be significantly improved by improving the reliability of reactor trip and passive decay heat removal system operation. The findings presented in this paper are expected to contribute toward future applications of PSA for assessing unconventional nuclear reactors in their conceptual design phases.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4181-4194
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• 2022

Main steam safety valves are commonly used in nuclear power plants to provide final protections from overpressure events. Blowdown and dynamic stability are two critical characteristics of safety valves. However, due to the parameter sensitivity and multi-parameter features of safety valves, using traditional method to design and/or optimize them is generally difficult and/or inefficient. To overcome these problems, a surrogate model-based valve design optimization is carried out in this study, of particular interest are methods of valve surrogate modeling, valve parameters global sensitivity analysis and valve performance optimization. To construct the surrogate model, Design of Experiments (DoE) and Computational Fluid Dynamics (CFD) simulations of the safety valve were performed successively, thereby an ensemble surrogate model (E-AHF) was built for valve blowdown and stability predictions. With the developed E-AHF model, global sensitivity analysis (GSA) on the valve parameters was performed, thereby five primary parameters that affect valve performance were identified. Finally, the k-sigma method is used to conduct the robust optimization on the valve. After optimization, the valve remains stable, the minimum blowdown of the safety valve is reduced greatly from 13.30% to 2.70%, and the corresponding variance is reduced from 1.04 to 0.65 as well, confirming the feasibility and effectiveness of the optimization method proposed in this paper.

• 농업생명과학연구
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• 제43권2호
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• pp.47-53
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• 2009

본 연구는 국화 재배온실에서 사용되고 있으면서 수작업 시 많은 노동시간이 요구되는 국화 유인네트를 대상으로 작업의 소요시간을 대폭 절감시켜 재배 작기를 증가시킴은 물론 국화재배용 유인네트의 베드를 기존 및 신축 국화재배 온실에 보급할 목적으로 네트의 베드를 설계한 후, 베드 및 기존 온실의 구조적 안전성을 검토하였다. 총 15종류의 부재에 대한 단면의 특성과 역학적 성질을 검토하고 각각의 부재에 대한 응력비를 검토한 결과, 이들 파이프 중 ${\phi}38.1{\times}1.7t$와 ${\phi}38.1{\times}2.0t$의 경우, 응력비는 초과하지 않았지만 처짐의 제한 값인 10mm를 초과하였으므로 응력비와 처짐의 제한 값을 모두 만족하는 최적 파이프는 ${\phi}48.1{\times}1.5t$인 것으로 나타났다. 유인네트의 베드를 중방에 설치할 경우, 베드 하중, 풍하중 및 설하중 때문에 중방은 트러스로 설계되어야 안전한 것으로 나타났다. 중방을 트러스로 시공할 경우에도 풍하중 및 설하중이 작용하면, 베드의 유무에 관계없이 거창지역의 국화재배 온실의 경우, 기둥과 방풍벽에 발생되는 응력은 허용응력을 초과하기 때문에 적절한 보강이 필요한 것으로 나타났다.

• 해양환경안전학회지
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• 제29권2호
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• pp.127-138
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• 2023

국제사회는 글로벌 기후위기 극복을 위해 2050년까지 탄소중립(Net Zero)을 목표로 다양한 탈탄소 에너지원 개발을 지속하고 있다. 우리 정부에서도 '재생에너지 3020' 정책을 수립하고 태양광이나 풍력을 이용한 에너지 개발계획을 추진함에 따라 해상 풍력발전단지와 같이 연안해역에서 기존에는 볼 수 없었던 대규모 해양개발사업이 추진되고 있다. 해양시설물은 선박의 입장에서 볼 때는 항행 장애물의 일종이며, 해양시설물 설치에 따라 좁아진 수역에서 선박 간 충돌사고 발생 또는 선박과 해양시설물의 접촉사고 발생시 환경오염 및 인명피해 등의 발생이 우려된다. 이에 국내외의 해상풍력발전단지 개발계획을 살펴보고 풍력단지에서 선박의 안전한 통항을 보장하기 위한 제도적 장치가 완비되어 있는지 분석하였으며, 해외의 입법 사례와 국내 법규를 비교하여 법적 사각지대를 해소하기 위한 새로운 법령안을 제안함으로써 대한민국의 관할해역에서 해양시설물의 안전한 운영과 선박의 안전한 통항을 기대하였다.

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.103-110
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• 2021

In the development of a Risk Monitor probabilistic safety assessment (PSA) model from the basic PSA model of a nuclear power plant, the modeling of common-cause failure (CCF) is very important. At present, some approximate modeling methods are widely used, but there lacks criterion of modeling accuracy and error analysis. In this paper, aiming at ensuring the accuracy of risk assessment and minimizing the Risk Monitor PSA models size, we present three basic issues of CCF model resulted from the changes of a nuclear power plant configuration, put forward corresponding modeling methods, and derive accuracy criteria of CCF modeling based on minimum cut sets and risk indicators according to the requirements of risk monitoring. Finally, a nuclear power plant Risk Monitor PSA model is taken as an example to demonstrate the effectiveness of the proposed modeling method and accuracy criteria, and the application scope of the idea of this paper is also discussed.

• Nuclear Engineering and Technology
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• 제41권2호
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• pp.163-170
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• 2009

In order to evaluate high-temperature oxidation behavior of the advanced alloy cladding under LOCA conditions, isothermal oxidation tests in steam were performed with cladding specimens prepared from high burnup PWR fuel rods that were irradiated up to 79 MWd/kg. Cladding materials were $M5^{(R)}$ and $ZIRLO^{TM}$, which are Nb-containing alloys. Ring-shaped specimens were isothermally oxidized in flowing steam at temperatures from 1173 to 1473 K for the duration between 120 and 4000s. Oxidation rates were evaluated from measured oxide layer thickness and weight gain. A protective effect of the preformed corrosion layer is seen for the shorter time range at the lower temperatures. The influence of pre-hydriding is not significant for the examined range. Alloy composition change generally has small influence on oxidation in the examined temperature range, though $M5^{(R)}$ shows an obviously smaller oxidation constant at 1273 K. Consequently, the oxidation rates of the high burnup $M5^{(R)}$ and $ZIRLO^{TM}$ cladding are comparable or lower than that of unirradiated Zircaloy-4 cladding.

• Nuclear Engineering and Technology
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• 제52권4호
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• pp.784-790
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• 2020

The accelerated corrosion by Flow Accelerated Corrosion (FAC) has caused unexpected rupture of piping, hindering the safety of nuclear power plants (NPPs) and sometimes causing personal injury. For the safety, it may be necessary to select some pipes in terms of condition monitoring and to measure the change in thickness of pipes in real time. Direct current potential drop (DCPD) method has advantages in on-line monitoring of pipe wall thinning. However, it has a disadvantage in that it is difficult to quantify thinning due to various thinning shapes and thus there is a limitation in application. The machine learning approach has advantages in that it can be easily applied because the machine can learn the signals of various thinning shapes and can identify the thinning using these. In this paper, finite element analysis (FEA) was performed by applying direct current to a carbon steel pipe and measuring the potential drop. The fundamental machine learning was carried out and the piping thinning model was developed. In this process, the features of DCPD to thinning were proposed.