• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2534-2546
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• 2021

Nuclear emergency preparedness and response is an essential part to ensure the safety of nuclear power plant (NPP). Key support technologies of nuclear emergency decision-making usually consist of accident diagnosis, source term estimation, accident consequence assessment, and protective action recommendation. Source term estimation is almost the most difficult part among them. For example, bad communication, incomplete information, as well as complicated accident scenario make it hard to determine the reactor status and estimate the source term timely in the Fukushima accident. Subsequently, it leads to the hard decision on how to take appropriate emergency response actions. Hence, this paper aims to develop a method for rapid source term estimation to support nuclear emergency decision making in pressurized water reactor NPP. The method aims to make our knowledge on NPP provide better support nuclear emergency. Firstly, this paper studies how to build a Bayesian network model for the NPP based on professional knowledge and engineering knowledge. This paper presents a method transforming the PRA model (event trees and fault trees) into a corresponding Bayesian network model. To solve the problem that some physical phenomena which are modeled as pivotal events in level 2 PRA, cannot find sensors associated directly with their occurrence, a weighted assignment approach based on expert assessment is proposed in this paper. Secondly, the monitoring data of NPP are provided to the Bayesian network model, the real-time status of pivotal events and initiating events can be determined based on the junction tree algorithm. Thirdly, since PRA knowledge can link the accident sequences to the possible release categories, the proposed method is capable to find the most likely release category for the candidate accidents scenarios, namely the source term. The probabilities of possible accident sequences and the source term are calculated. Finally, the prototype software is checked against several sets of accident scenario data which are generated by the simulator of AP1000-NPP, including large loss of coolant accident, loss of main feedwater, main steam line break, and steam generator tube rupture. The results show that the proposed method for rapid source term estimation under nuclear emergency decision making is promising.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.36 no.6
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• pp.655-662
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• 2012

Gas leakage in an oil refinery causes damage to the environment and unsafe conditions. Therefore, it is necessary to develop a technique that is able to detect the location of the leakage and to filter abnormal gas-leakage signals from normal background noise. In this study, the adaptation filter of the finite impulse response (FIR) least mean squares (LMS) algorithm and a cross-correlation function were used to develop a leakage-predicting program based on LABVIEW. Nitrogen gas at a high pressure of 120 kg/$cm^2$ and the assembled equipment were used to perform experiments in a reverberant chamber. Analysis of the data from the experiments performed with various hole sizes, pressures, distances, and frequencies indicated that the background noise occurred primarily at less than 1 kHz and that the leakage signal appeared in a high-frequency region of around 16 kHz. Measurement of the noise sources in an actual oil refinery revealed that the noise frequencies of pumps and compressors, which are two typical background noise sources in a petrochemical plant, were 2 kHz and 4.5 kHz, respectively. The fact that these two signals were separated clearly made it possible to distinguish leakage signals from background noises and, in addition, to detect the location of the leakage.

• Journal of Energy Engineering
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• v.27 no.2
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• pp.14-25
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• 2018

The combined cycle power plant has a cycle of operating the gas turbine with fuel, such as natural gas, and then producing steam using residual heat. The fuel gas is supplied to the gas turbine at a level of 4 to 5 MPa, $200^{\circ}C$ through a compressor and a heat exchanger. In this study, the risk assessment method considering the piping system stress was carried out for safe operation and soundness of the gas fuel supply piping system. The API 580/581 RBI code, which is well known for its risk assessment techniques, is limited to reflect the effect of piping stress on risk. Therefore, the systematic stress of the pipeline is analyzed by using the piping analysis. For the study, the piping system stress analysis was performed using design data of a gas fuel supply piping of a combined cycle power plant. The result of probability of failure evaluated by the API code is compared to the result of stress ratio by piping analysis.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.3
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• pp.389-396
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• 2018

The Kori-1 Nuclear Power Plant (NPP), WH 2-Loop Pressurized Water Reactor (PWR) operated for approximately 40 years in Korea, was permanently ceased on June 18, 2017. To reduce worker exposure to radiation by reducing the dose rate in the system before starting main decommissioning activities, the permanently ceased Kori-1 NPP will be subjected to full system decontamination. Generally, the range of system decontamination includes Reactor Pressure Vessels (RPV), Pressurizer (PZR), Steam Generators (SG), Chemical & Volume Control System (CVCS), Residual Heat Removal System (RHRS), and Reactor Coolant System (RCS) piping. In order to decontaminate these systems and equipment in an effective manner, it is necessary to evaluate the influence of the flow characteristics in the RCS during the decontamination period. There are various methods of providing circulating flow rate to the system decontamination. In this paper, the flow characteristics in Kori-1 NPP reactor coolant according to RHR pump operation were evaluated. The evaluation results showed that system decontamination using an RHR pump was not effective at decontamination due first to impurities deposited in piping and equipment, and second to the extreme flow unbalance in the RCS caused deposition of impurities.

• Korean Journal of Medicinal Crop Science
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• v.11 no.5
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• pp.352-357
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• 2003

Volatile flavor compounds Angelica gigas Nakai and Angelica acutiloba Kitagawa were extracted by SDE (simultaneous steam distillation & extraction) using the mixture of n-pentane and diethylether (1:1, v/v) as an extract solvent and analyzed by GC-FID and GC-MS. The amount of essential oils of top part and root in Angelica gigas were obtained in 0.063% (v/w) and 0.389% (v/w) yields as a fresh weight base, respectively. The main compounds in top parts and the root were identified as nonane (7.51% and 24.49%, respectively), ${\alpha}-pinene$ (14.64% and 31.75%), limonene+${\beta}-phellandrene$ (14.01% and 9.66%), ${\gamma}-terpinene$ (7.85% and 1.20%), germacrene-d (5.85% and 0.22%), (E,E)-${\alpha}-farnesene$ (6.05% and 1.40%), ${\beta}-eudesmol$ (5.26% and 1.84%). Although these compounds were present in both parts. The results showed large differences in. the concentrations of them much varied. The amount of essential oils stem and leaf obtained (0.068% and 0.127% in A. gigas) and (0.153% and 0.243% in A. acutiloba) yields as a fresh weight base, respectively. More than 18 and 32 components in stem and leaf have been identified, which of main components in A. gigas were ${\alpha}-pinene$, myrcene, limonene, germacrene-d, eudesmol and butylphthalide, but germacrene-d and butylphthalide contents were also different in stem and leaf. And more than 21 and 32 components in A. acutiloba were ${\gamma}-terpinene$ and butylphthalide. Volatile compounds were very different in both species.

• Journal of Advanced Marine Engineering and Technology
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• v.41 no.3
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• pp.182-190
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• 2017

All liquids contain a small amount of gaseous components and the amount of gases dissolved in a liquid is in accordance with Henry's Law. In a multi-stage thermal-type seawater desalination plant, as the supplied seawater undergoes variations in temperature and pressure in each evaporator, the gases dissolved in the seawater are discharged from the liquid. The discharged gases are carbon dioxide, nitrogen, oxygen, and argon, and these emitted gases are non-condensable. From the viewpoint of convective heat transfer, the evaluation of non-condensable gas released during a vacuum evaporation process is a very important design factor because the non-condensable gases degrade the performance of the cooler. Furthermore, in a thermal-type seawater desalination plant, most evaporators operate under vacuum, which maintained through vacuum system such as a steam ejector or a vacuum pump. Therefore, for the proper design of a vacuum system, estimating the non-condensable gases released from seawater is highly crucial. In the study, non-condensable gases released in a thermal-type seawater desalination plant were calculated quantitatively. The calculation results showed that the NCG releasing rate decreased as the stage comes getting a downstream and it was proportional to the freshwater production rate.

• Journal of the Korean Society for Nondestructive Testing
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• v.36 no.6
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• pp.504-509
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• 2016

Defects in dissimilar metal welds are reported to be on the increase during the operating lifespan and aging of nuclear power plants. In Korea, reported cases of defects due to dissimilar metal welds include the drain nozzle of a steam generator and RCS hot tube sampling nozzles. Therefore, there is an urgent need to develop a reliable automated nondestructive inspection technique and a system for the inspection of dissimilar metal welds of small diameter nozzles in a high radiation area of a nuclear power plant. In this study, to ensure effective defect inspection of small diameter nozzles (RCS high-temperature tube sampling nozzle) of a nuclear power plant, three different methods were developed. These include: (1) optimum inspection probe design by beam simulation, (2) multi-directions UT optimum inspection technique for the inspection of small diameters of different welded parts, and (3) remote control automatic inspection system. The developed technique and systems have been verified to be suitable for use in the inspection of defects in smaller diameter nozzles in nuclear power plants.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.575-602
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• 2009

In the wake of the Three Mile Island accident, vigorous research efforts were initiated to acquire a basic knowledge of the progression and consequences of accidents that involve a substantial degree of core degradation and melting. The primary emphasis of this research was placed on containment integrity, with: i) hydrogen combustion-detonation, ii) steam explosion, iii) direct containment heating (DCH), and iv) melt attack on the BWR Mark-I containment shell identified as energetic processes that could lead to early containment failure (i.e., within the first 24 hours of the accident). Should the core melt fail the reactor vessel, then non-condensable gas production from Molten Core-Concrete Interaction (MCCI) was identified as a mechanism that could fail the containment by pressurization over the long term. One signification question that arose as part of this investigation was the effectiveness of water in terminating an MCCI by flooding the interacting masses from above, thereby quenching the molten core debris and rendering it permanently coolable. Successful quenching of the core melt would prevent basemat melt through, as well as continued containment pressurization by non-condensable gas production, and so the accident progression would be successfully terminated without release of radioactivity to the environment. Based on these potential merits, ex-vessel corium coolability has been the focus of extensive research over the last 20 years as a potential accident management strategy for current plants. In addition, outcomes from this research have impacted the accident management strategies for the Gen III+LWR plant designs that are currently being deployed around the world. This paper provides: i) an historical overview of corium coolability research, ii) summarizes the current status of research in this area, and iii) highlights trends in severe accident management strategies that have evolved based on the findings from this work.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.21 no.1
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• pp.22-31
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• 1997

The grain boundary etching method as a method for assessing degradation of structural materials has received much attention because it is simple, inexpensive and easy to apply to real components. In this study, the effectiveness of the method is verified by successfully applying the technique to in-service components of aged fossil power plants such as main steam pipes, boiler headers an turbine rotors. A new degradation parameter, intersecting number ratio (N$_{1}$/N$_{0}$), is employed. The intersecting number ratio (N$_{1}$/N$_{0}$) is defined as the ratio of intersection number (N$_{1}$) obtained from 5-minute picric acid etched surface to the number (N$_{0}$) obtained from nital etched surface. Two kinds of test materials, 2.25Cr-1Mo steel and 1Cr-1Mo-0.25V steel, were artificially thermal-aged at 630.deg. C in different levels of degradation., (N$_{1}$/N$_{0}$) were measured. And, correlations between the measured values and LMP values calculated from aging temperature and aging time were sought. To check the validity of the correlations obtained in laboratory, similar data were measured from service components in four old Korean fossil power plants. These on-site measurement data were in good correlation with those obtained in the laboratory.oratory.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.12
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• pp.1265-1273
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• 2015

A coal handling machine is a type of equipment used for loading coal, the main material in a steam power plant, along a conveyer belt from a ship, and is placed after the driving chain bucket. However, studies on the boom hoisting cylinder, which is a hydraulic system used to control the angle of the boom based on loading location, indicate that domestic models are insufficient, and are thereby often substituted with a foreign product. In this study, a technique for analyzing the contact pressure in a thick-walled cylinder was established by comparing the contact pressure, which is calculated theoretically based on the results obtained from FEM simulation, and by checking whether the working oil is leaking from the boom hoisting cylinder using a v-seal. In addition, the driving motion was simulated according to the strokes of the cylinder, and the structural stability was verified under the maximum output conditions.