• Nuclear Engineering and Technology
• /
• 제44권3호
• /
• pp.249-260
• /
• 2012

Main focus of the combined nuclear research activities at Aachen University (RWTH) and the Research Center J$\ddot{u}$lich (J$\ddot{U}$LICH) is the experimental and analytical investigation of containment phenomena and processes. We are deeply convinced that reliable simulations for operation, design basis and beyond-design basis accidents of nuclear power plants need the application of so-called lumped-parameter (LP) based codes as well as computational fluid dynamics (CFD) codes in an indispensable manner. The LP code being used at our institutions is the GRS code COCOSYS and the CFD tool is ANSYS CFX mostly used in German nuclear research. Both codes are applied for safety analyses especially of beyond design accidents. Focal point of the work is containment thermal-hydraulics, but source term relevant investigations for aerosol and iodine behavior are performed as well. To increase the capability of COCOSYS and CFX detailed models for specific features, e.g. recombiner behavior including chimney effect, building condenser, and wall condensation are developed and validated against facilities at different scales. The close connection between analytical and experimental activities is notable and identifying feature of the RWTH/J$\ddot{U}$LICH activities.

• Nuclear Engineering and Technology
• /
• 제38권8호
• /
• pp.697-732
• /
• 2006

In this paper, a historical review of the developments in the safety of LWR power plants is presented. The paper reviews the developments prior to the TMI-2 accident, i.e. the concept of the defense in depth, the design basis, the large LOCA technical controversies and the LWR safety research programs. The TMI-2 accident, which became a turning point in the history of the development of nuclear power is described briefly. The Chernobyl accident, which terrified the world and almost completely curtailed the development of nuclear power is also described briefly. The great international effort of research in the LWR design-base and severe accidents, which was, respectively, conducted prior to and following the TMI-2 and Chernobyl accidents is described next. We conclude that with the knowledge gained and the improvements in plant organisation/management and in the training of the staff at the presently-installed nuclear power stations, the LWR plants have achieved very high standards of safety and performance. The Generation 3+LWR power plants, next to be installed, may claim to have reached the goal of assuring the safety of the public to a very large extent. This review is based on the historical developments in LWR safety that occurred primarily in USA, however, they are valid for the rest of the Western World. This review can not do justice to the many fine contributions that have been made over the last fifty years to the cause of LWR safety. We apologize if we have not mentioned them. We also apologize for not providing references to many of the fine investigations, which have contributed towards LWR safety earning the conclusions that we describe just above.

• 한국안전학회지
• /
• 제27권4호
• /
• pp.90-95
• /
• 2012

To prevent similar accidents with the basis of industrial accidents already occurred in industrial plants, it would be possible only after true causes are grasped. Unfortunately, however, most accident investigation carried out with the basis of legal regulation failed to grasp them so that similar accidents have been repeated without cease. This research aimed to find out differences between results from conventional accident investigation and those from human error analysis, and to draw out effective and practical counter-plans against industrial accidents occurred repeatedly in an autoglass manufacturing company. As for analysis, about 110 accident cases that occurred for last 7 years were collected, and by adopting the Comprehensive Human Error Analysis Technique developed by the previous researchers, not direct causes but basic fundamental causes that might induce workers to human errors were sought. In consequence, the result showed that facility factors or environmental factors such as improper layout, mistakes in engineering design, and malfunction of interlock system were authentic major accident causes as opposed to managerial factors such as personal carelessness or failure to wearing personal protective equipments, and/or improper work methods.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 2002년도 추계학술발표회요약집
• /
• pp.206-206
• /
• 2002

• 한국압력기기공학회 논문집
• /
• 제15권2호
• /
• pp.64-70
• /
• 2019

In recent years, attention has been paid to the integrity of steam generator (SG) tubes due to severe accident and beyond design basis accident conditions. In these transient conditions, steam generator tubes may be damaged by high temperature and pressure, which might result in a risk of fission products being released to the environment due to the failure. Alloy 690 which has increased the Cr content has been replaced for the SG tube due to its high corrosion resistance against stress corrosion cracking (SCC). However, there is lack of research on the high temperature creep rupture and life prediction model of Alloy 690. In this study, creep test was performed to estimate the high temperature creep rupture life of Alloy 690 using tube specimens. Based on manufacturer's creep data and creep test results performed in this study, creep life prediction was carried out using the Larson-Miller (LM) Parameter, Orr-Sherby-Dorn (OSD) parameter, Manson-Haford (MH) parameter, and Wilshire's approach. And a hyperbolic sine (sinh) function to determine master curves in LM, OSD and MH parameter methods was used for improving the creep life estimation of Alloy 690 material.

• 한국안전학회지
• /
• 제35권5호
• /
• pp.121-127
• /
• 2020

Since the Fukushima nuclear accident in 2011, the public were concerned about the safety of Nuclear Power Plants (NPPs) in extreme natural disaster situations, such as earthquakes, flooding, heavy rain and tsunami, have been increasing around the world. Accordingly, the Stress Test was conducted in Europe, Japan, Russia, and other countries by reassessing the safety and response capabilities of NPPs in extreme natural disaster situations that exceed the design basis. The extreme natural disaster can put the NPPs in beyond-design-basis conditions such as the loss of the power system and the ultimate heat sink. The behaviors and capabilities of NPPs with losing their essential safety functions should be measured to find and supplement weak areas in hardware, procedures and coping strategies. The Loss of Ultimate Heat Sink (LUHS) accident assumes impairment of the essential service water system accompanying the failure of the component cooling water system. In such conditions, residual heat removal and cooling of safety-relevant components are not possible for a long period of time. It is therefore very important to establish coping strategies considering all available equipment to mitigate the consequence of the LUHS accident and keep the NPPs safe. In this study, thermal hydraulic behavior of the LUHS event was analyzed using RELAP5/Mod3.3 code. We also performed the sensitivity analysis to identify the effects of the operator recovery actions and operation strategy for charging pumps on the results of the LUHS accident.

• 한국압력기기공학회 논문집
• /
• 제18권1호
• /
• pp.19-25
• /
• 2022

Nuclear Power Plants (NPP) should be designed to have sufficient safety margins and to ensure seismic safety against earthquake that may occur during the plant life time. After the 9.12 Gyeongju earthquake accident, the structural integrity of nuclear power plants due to the beyond design basis earthquake is one of key safety issues. Accordingly, it is necessary to conduct structural integrity evaluations for domestic NPPs under beyond design basis earthquake. In this study, the Level 3 LBB (Leak Before Break) evaluation was performed by considering the beyond design basis earthquake for the surge line of a OPR1000 plant of which design basis earthquake was set to be 0.2g. The beyond design basis earthquake corresponding to peak ground acceleration 0.4g at the maximum stress point of the surge line was considered. It was confirmed that the moment behaviors of the hot leg and pressurized surge nozzle were lower than the maximum allowable loading in moment-rotation curve. It was also confirmed that the LBB margin could be secured by comparing the LBB margin through the Level 2 method. It was judged that the margin was secured by reducing the load generated through the compliance of the pipe.

• 대한전기학회:학술대회논문집
• /
• 대한전기학회 2006년 학술대회 논문집 정보 및 제어부문
• /
• pp.600-602
• /
• 2006

Some abnormal signals diagnostics and analysis through an important equivalent circuits modeling for passive elements under severe accident conditions have been performed. Unlike the design basis accidents, there are inherently some uncertainties in the instrumentation capabilities under the accident conditions. So, the circuit simulation analysis and diagnosis methods are used to assess instruments in detail when they give apparently abnormal readings as an accident alternative method. The simulations can be useful to investigate what the signal and circuit characteristics would be when similar to a variety of symptoms that can result from the environmental conditions such as high temperature, humidity, and pressure condition. In this paper, a new simulator through an analysis of the important equivalent circuits modeling under temperature accident conditions has been designed, the designed simulator is composed of the LabVIEW code as a main tool and the out-put file of the Multi-SIM code as an engine tool is exported to in-put file of the LabVIEW code. The procedure for the simulator design was divided into two design steps, of which the first step was the diagnosis method, the second step was the circuit simulator for the signal processing tool. It has three main functions which are a signal processing tool, an accident management tool, and an additional guide from the initial screen. This simulator should be possible that it could be applied a output signal analysis to some transient signal by variation of the resistance passive elements in the R-L-C equivalent circuit modeling under various degraded conditions in NPPs.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1999년도 춘계학술발표회요약집
• /
• pp.185-185
• /
• 1999

• Nuclear Engineering and Technology
• /
• 제52권5호
• /
• pp.937-946
• /
• 2020

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.