• Nuclear Engineering and Technology
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• 제51권1호
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• pp.104-115
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• 2019

Since the Fukushima Daiichi accident, there has been an emphasis on the risk resulting from multi-unit accidents. Human reliability analysis (HRA) is one of the important issues in multi-unit probabilistic safety assessment (MUPSA). Hence, there is a need to properly identify all the human and organizational factors relevant to a multi-unit incident scenario in a nuclear power plant (NPP). This study identifies and categorizes the human and organizational factors relevant to a multi-unit incident scenario of NPPs based on a review of relevant literature. These factors are then analyzed to ascertain all possible unit-to-unit interactions that need to be considered in the multi-unit HRA and the pattern of interactions. The human and organizational factors are classified into five categories: organization, work device, task, performance shaping factors, and environmental factors. The identification and classification of these factors will significantly contribute to the development of adequate strategies and guidelines for managing multi-unit accidents. This study is a necessary initial step in developing an effective HRA method for multiple NPP units in a site.

• Nuclear Engineering and Technology
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• 제53권3호
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• pp.793-803
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• 2021

The small PWR has been paid more and more attention due to its diversity of application and flexibility in the site selection. However, the large core power density, the small containment space and the rapid accident progress characteristics make it difficult to control the containment pressure like the traditional PWR during the LOCA. The pressure suppression system has been used by the BWR since the early design, which is a suitable technique that can be applied to the small PWR. Since the configuration and operating conditions are different from the BWR, the pressure suppression system should be redesigned for the small PWR. Conducting the experiments on the scale down test facility is a good choice to reproduce the prototypical phenomena in the test facility, which is both economical and reasonable. A systematic scaling method referring to the H2TS method was proposed to determine the geometrical and thermohydraulic parameters of the pressure suppression containment response test facility for the small PWR conceptual design. The containment and the pressure suppression system related thermohydraulic phenomena were analyzed with top-down and bottom-up scaling methods. A set of the scaling criteria were obtained, through which the main parameters of the test facility can be determined.

• 시스템엔지니어링학술지
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• 제16권2호
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• pp.67-77
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• 2020

Traditionally nuclear thermal hydraulic and nuclear safety has relied on numerical simulations to predict the system response of a nuclear power plant either under normal operation or accident condition. However, this approach may sometimes be rather time consuming particularly for design and optimization problems. To expedite the decision-making process data-driven models can be used to deduce the statistical relationships between inputs and outputs rather than solving physics-based models. Compared to the traditional approach, data driven models can provide a fast and cost-effective framework to predict the behavior of highly complex and non-linear systems where otherwise great computational efforts would be required. The objective of this work is to develop an AI algorithm to predict the peak fuel cladding temperature as a metric for the successful implementation of FLEX strategies under extended station black out. To achieve this, the model requires to be conditioned using pre-existing database created using the thermal-hydraulic analysis code, MARS-KS. In the development stage, the model hyper-parameters are tuned and optimized using the talos tool.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2836-2846
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• 2020

During severe nuclear power plant (NPP) accidents, a H₂/CO mixture can be generated in the reactor pressure vessel by core degradation and in the containment as well by molten corium-concrete interaction. In spite of its importance, a state-of-the-art methodology predicting H₂/CO combustion risk relies predominantly on empirical correlations. It is therefore necessary to develop a proper methodology for flammability evaluation of H₂/CO mixtures at ex-vessel phases characterized by three factors: CO concentration, high temperature, and diluents. The developed methodology adopted Le Chatelier's law and a calculated non-adiabatic flame temperature model. The methodology allows the consideration of the individual effect of the heat transfer characteristics of hydrogen and carbon monoxide on low flammability limit prediction. The accuracy of the developed model was verified using experimental data relevant to ex-vessel phase conditions. With the developed model, the prediction accuracy was improved substantially such that the maximum relative prediction error was approximately 25% while the existing methodology showed a 76% error. The developed methodology is expected to be applicable for flammability evaluation in chemical as well as NPP industries.

• Nuclear Engineering and Technology
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• 제11권3호
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• pp.195-202
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• 1979

In this paper, a cause-consequence chart is applied to evaluate the probability that the containment spray system in a nuclear power plant may not be woring properly, given a demand for spryaing at loss of coolant accident (LOCA). It is shown how the diagram provides a basis for calculating two probability measures for malfunctioning of this system, in case the test policy of the system is taken into account, i.e., average probability that the containment spray cannot be established, and average probability that the containment spray is established : spray stops before the required operating time is over.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.475-479
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• 1995

During the hypothetical loss of coolant accident in the nuclear power plant the emergency core cooling water could not penetrate to the reactor core when the steam flow rate from the reactor core exceeds CCFL (Countercurrent flow limitation). The CCFL generated by earlier investigators are developed under the steady gas flow. However the flow instability in the reactor loop could generate oscillatory steam flow, hence their applicability under oscillating flow should be investigated. In this work, an experimental investigation of countercurrent flow in the vertical flow channel has been conducted under oscillatory gas flow. Pulsation of gas under oscillatory flow disturbs the flow pattern significantly and prevents flooding (CCFL) when its minimum value is less than the threshold gas flow rate value.

• Nuclear Engineering and Technology
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• 제46권6호
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• pp.825-836
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• 2014

Predicting radioactive fission product (FP) behaviors in the reactor coolant system and the containment of a nuclear power plant (NPP) is one of the major concerns in the field of reactor safety, since the amount of radioactive FP released into the environment during the postulated accident sequences is one of the major regulatory issues. Radioactive FPs circulating in the primary coolant loop and released into the containment are basically in the form of gas or aerosol. In this study, a multi-component and multi-sectional analysis module for aerosol fission products has been developed based on the MAEROS model [1,2], and the aerosol transport model has been developed and verified against an analytic solution. The deposition of aerosol FPs to the surrounding structural surfaces is modeled with recent research achievements. The developed aerosol analysis model has been successfully validated against the STORM SR-11 experimental data [3], which is International Standard Problem No. 40. Future studies include the development of the resuspension, growth, and chemical reaction models of aerosol fission products.

• Nuclear Engineering and Technology
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• 제50권3호
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• pp.526-530
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• 2018

Human fingernails were used to estimate the radiation dose via electron paramagnetic resonance measurements of radiation-induced radicals. The limiting factors in this research were mechanically induced electron paramagnetic resonance signals due to the mechanical stress during the preparation of the samples. Therefore, different treatment methods of fingernails were used to reduce the mechanically induced signals. The results demonstrate that the mechanically induced and radiation-induced signals have apparently different microwave power saturation behaviors. In addition, the mechanically induced signal shows a fading evolution over time and reaches a constant value. Chemical treatment using the different reagents showed that the minimum mechanically induced signal was obtained using the dithiothreitol reagent. The dose-response curves of the samples treated with dithiothreitol for 30 minutes demonstrated a greater linearity than those of samples treated for 5 minutes. Therefore, to find an unknown absorbed dose in a fingernail sample using a calibration curve, we recommend adopting the mentioned chemical treatment procedure to reduce the uncertainty.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2069-2087
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• 2023

In nuclear power plant (NPP) accidents, the containment is subject to high temperatures and high internal pressures, which may further trigger serious chain accidents such as core meltdown and hydrogen explosion, resulting in a significantly higher accident level. Therefore, studying the mechanical performance of a containment under high temperature and high internal pressure is relevant to the safety of NPPs. Based on similarity principles, the 1:3.2 scale model of a prestressed concrete containment vessel (PCCV) of a NPP was designed. The loading method, which considers the thermal-pressure coupling conditions, was used. The mechanical response of the PCCV was investigated with a simultaneous increase in internal pressure and temperature, and the failure mechanism of the PCCV under thermal-pressure coupling conditions was revealed.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2039-2049
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• 2024

The radioactive pollutant could migrate to the downstream urban area under the action of atmospheric dispersion due to the turbulent mixing under actual pollution accidents. A scenario in which radioactive contaminants from the upstream (for example, a nearshore nuclear power plant accident) migrates to the downstream urban blocks have been considered in this study. Numerical simulations using computational fluid dynamics (CFD) are then conducted to investigate the effects of the urban morphology (building packing density and layout) on the atmospheric dispersion of radioactive pollutants in this scenario. The building packing density and structure can significantly affect urban areas' mean flow pattern and the turbulent kinetic energy (TKE). The flow pattern and the TKE distribution influence the radioactive pollution dispersion. It is found that the radioactive pollution at the urban canyons is significantly affected by the vertical transport at the canyon. A comparison of the distributions of radioactive and traditional non-radioactive pollutants is also provided.