• Nuclear Engineering and Technology
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• 제55권8호
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• pp.3017-3029
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• 2023

In the late in-vessel phase of a nuclear reactor severe accident, the internal heat transfer and crust evolution during the debris bed melting process have important effects on the thermal load distribution along the vessel wall, and further affect the reactor pressure vessel (RPV) failure mode and the state of melt during leakage. This study coupled the phase change model and large eddy simulation to investigate the variations of the temperature, melt liquid fraction, crust and heat flux distributions during the debris bed melting process in the hypothetical severe accident of HPR1000. The results indicated that the heat flow towards the vessel wall and upper surface were similar at the beginning stage of debris melting, but the upward heat flow increased significantly as the development of the molten pool. The maximum heat flux towards the vessel wall reached 0.4 MW/m². The thickness of lower crust decreased as the debris melting. It was much thicker at the bottom region with the azimuthal angle below 20° and decreased rapidly at the azimuthal angle around 20-50°. The maximum and minimum thicknesses were 2 and 90 mm, respectively. By contrast, the distribution of upper crust was uniform and reached stable state much earlier than the lower crust, with the thickness of about 10 mm. Moreover, the sensitivity analysis of initial condition indicated that as the decrease of time interval from reactor scram to debris bed dried-out, the maximum debris temperature and melt fraction became larger, the lower crust thickness became thinner, but the upper crust had no significant change. The sensitivity analysis of in-vessel retention (IVR) strategies indicated that the passive and active external reactor vessel cooling (ERVC) had little effect on the internal heat transfer and crust evolution. In the case not considering the internal reactor vessel cooling (IRVC), the upper crust was not obvious.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4514-4521
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• 2022

In this study, the effects of heat and radiation on the degradation behaviour of fluoroelastomer under simulated normal operation and a severe accident environment were investigated using sequential testing of gamma irradiation and thermal degradation. Tensile properties and Shore A hardness were measured, and thermogravimetric analysis was used to evaluate the degradation behaviour of fluoroelastomer. Fourier transform infrared spectroscopy and X-ray photoelectron spectroscopy were used to characterize the structural changes of the fluoroelastomer. Heat and radiation generated in nuclear power plant break and deform the chemical bonds, and fluoroelastomer exposed to these environments have decreased C-H and functional groups that contain oxygen and double bonds such as C-O, C=O and C=C were generated. These functional groups were formed by auto oxidation by reacting free radicals generated from the cleaved bond with oxygen in the atmosphere. In this auto oxidation reaction, crosslinks were generated where bonded to each other, and the mobility of molecules was decreased, and as a result, the fluoroelastomer was hardened. This hardening behaviour occurred more significantly in the severe accident environment than in the normal operation condition, and it was found that thermal stability decreased with the generation of unstable structures by crosslinking.

• Nuclear Engineering and Technology
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• 제51권3호
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• pp.676-682
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• 2019

After the Fukushima Daiichi accident, there has been an increasing preference for passive safety features in the nuclear power industry. Some passive safety systems require limited active components to trigger subsequent passive operation. Under very serious accident conditions, passive safety features could be rendered inoperable or damaged. This study evaluates (i) the performance and effectiveness of the passive safety features of iPOWER (innovative Power Reactor), and (ii) whether a severe accident condition could be reached if the passive safety systems are damaged, namely the case of heat exchanger tube rupture. Analysis results show that the reactor coolant system remains in the hot shutdown condition without operator actions or electricity for over 72 h when the passive auxiliary feedwater systems (PAFSs) are operable without damage. However, heat exchanger tube rupture in the PAFS leads to core damage after about 18 h. Such results demonstrate that, to enhance the safety of iPOWER, maintaining the integrity of the PAFS is critical, and therefore additional protections for PAFS are necessary. To improve the reliability of iPOWER, additional battery sets are necessary for the passive safety systems using limited active components for accident mitigation under such extreme circumstances.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4522-4533
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• 2022

For severe accident assessment of reactor pressure vessel (RPV), it is important to develop an accurate model that can predict transient thermo-mechanical behavior of the RPV lower head under the given condition. The present study revisits the lower head failure with two- and three-dimensional finite element models. In particular, we aim to give clear insight regarding the effect of the three-dimensionality present in the distribution of the thickness and thermal load of the lower head. For a rigorous validation of the result, both the OLHF-1 and the OLHF-2 tests are considered in this study. The result suggests that the three-dimensional effect is not negligible as far as the failure location is concerned. The non-uniformity of the thickness distribution is found to affect the failure location and time. The thermal load, which may not be axisymmetric in general, has the most significant effect on the failure assessment. We also observe that the creep property can affect the global deformation of the lower head, depending on the applied mechanical load.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4134-4145
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• 2023

This paper proposes strain-based failure model of A533B1 pressure vessel steel to simulate failure, followed by application to OECD lower head failure (OLHF) test simulation for experimental validation. The proposed strain-based failure model uses simple constant and linear functions based on physical failure modes with the critical strain value determined either using the lower bound of true fracture strain or using the average value of total elongation depending on the temperature. Application to OECD Lower Head Failure (OLHF) tests shows that progressive deformation, failure time and failure location can be well predicted.

• Nuclear Engineering and Technology
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• 제44권3호
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• pp.237-248
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• 2012

Corium behavior in the lower plenum of the reactor vessel during a severe accident is very important, as this affects a failure mechanism of the lower head vessel and a thermal load to the outer reactor vessel under the IVR-ERVC (In-Vessel corium Retention through External Reactor Vessel Cooling) condition. This paper discusses the state of the art and technical issues on corium behavior in the lower plenum, such as initial corium pool formation characteristics and its transient behavior, natural convection heat transfer in various geometries, natural convection heat transfer with a phase change of melting and solidification, and corium interaction with a lower head vessel including penetrations of the ICI (In-Core Instrumentation) nozzle are discussed. It is recommended that more detailed analysis and experiments are necessary to solve the uncertainties of corium behavior in the lower plenum of the reactor vessel.

• 대한기계학회논문집A
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• 제24권11호
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• pp.2866-2874
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• 2000

This paper is concerned with the optimum design for external reinforcement of a nuclear reactor pressure vessel(RPV) in a severe accident. During the severe reactor accident of molten core, the temperature and the pressure in the nuclear reactor rise to a certain level depending on the initial and subsequent condition of a severs accident. The reis of the temperature and the internal pressure cause deterioration of the load carrying capacity and could cause failure of the RPV lower head. The deterioration of failure can be mitigated by the external cooling or the reinforcement of the lower head with additional structures. While the external cooling forces the temperature of an RPV to drop to the desired level, the reinforcement of the lower head can attain both the increase of the load carrying capacity and the temperature drop. The reinforcement of the lower head can be optimized to have the maximum effect on the collapse pressure and the temperature at the inner wall. Optimization results are compared to both the result without the reinforcement and the result with the designated reinforcement.

• 대한인간공학회지
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• 제35권3호
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• pp.135-142
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• 2016

Objective: Objective of this study is to provide characteristics of injury frequency and severity by driving condition in large truck-related traffic collisions. Background: Traffic accidents involving large trucks draw a lot of attention in accident prevention and management policies since they bring about severe human and financial damages. Method: In order to identify the major risk factors of accidents by driving condition, 255 recognized traffic accidents by large truck drivers were analyzed in terms of time of the day, road type, and shape of the road. Results: The driving conditions in the results are represented by the following form of combination, "Road Type (Non-expressway or Express) - Shape of Roads (Straight, Curved, Downhill, or Intersection) - Time of Accidents (Day or Night)". In the analysis of injury frequency, Non-expressway-Straight-Day condition was the most frequent one. Meanwhile, Expressway-Curved-Day, Non-expressway-Curved-Night and Non-expressway-Intersection-Night were evaluated as high level in view of injury severity. Also, Expressway-Straight-Night is the driving condition that is the highest in risk among the conditions that have to be managed as grade "High". Non-expressway-Straight-Night, Non-expressway-Downhill-Day, and Non-expressway-Curved-Day are also categorized as grade "High". Conclusion and Application: Safety managers in the fields require basic information on accident prevention that can be easily understood. The research findings will serve as a practical guideline for establishing preventive measures for traffic accidents.

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

• 한국소음진동공학회:학술대회논문집
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• 한국소음진동공학회 2001년도 추계학술대회논문집 II
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• pp.766-771
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• 2001

The package used to transport radioactive materials, which is called by cask, must be safe under normal and hypothetical accident conditions. These requirements for the cask design must be verified through test or finite element analysis. Since the cost for FE analysis is less than one for test. the verification by FE analysis is mainly used. But due to the complexity of mechanical behaviors. the results depends on how users apply the codes and it can cause severe errors during analysis. In this paper, finite element analysis is carried out for the 9 meters free drop and the puncture condition of the hypothetical accident conditions using LS-DYNA3D and ABAQUS/Explicit. We have investigated the analyzing technique for the free drop impact test of the cask and found several vulnerable cases to errors. The analyzed results were compared with each other. We have suggested a reliable and relatively simple analysis technique for the drop test of spent nuclear fuel casks.