• Nuclear Engineering and Technology
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• v.41 no.3
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• pp.327-334
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• 2009

Elastic wave is one of the most useful tools for non-destructive tests in nuclear power plants. Since the elastic properties are indispensable for analyzing the behaviors of elastic waves, they should be predetermined within an acceptable accuracy. Nuclear power plants are exposed to harsh environmental conditions and hence the structures are degraded. It means that the Young's modulus becomes unreliable and in-situ measurement of Young's modulus is required from an engineering point of view. Young's modulus is estimated from the group velocity of propagating waves. Because the flexural wave of a plate is inherently dispersive, the group velocity is not clearly evaluated in temporal signal analysis. In order to overcome such ambiguity in estimation of group velocity, Wigner-Ville distribution as the time-frequency analysis technique was proposed and utilized. To verify the proposed method, experiments for steel and acryl plates were performed with accelerometers. The results show good estimation of the Young's modulus of two plates.

• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.849-858
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• 2009

Risk caused by safety-critical instrumentation and control (I&C) systems considerably affects overall plant risk. As digitalization of safety-critical systems in nuclear power plants progresses, a risk model of a digitalized safety system is required and must be included in a plant safety model in order to assess this risk effect on the plant. Unique features of a digital system cause some challenges in risk modeling. This article aims at providing an overview of the issues related to the development of a static fault-tree-based risk model. We categorize the complicated issues of digital system probabilistic risk assessment (PRA) into four groups based on their characteristics: hardware module issues, software issues, system issues, and safety function issues. Quantification of the effect of these issues dominates the quality of a developed risk model. Recent research activities for addressing various issues, such as the modeling framework of a software-based system, the software failure probability and the fault coverage of a self monitoring mechanism, are discussed. Although these issues are interrelated and affect each other, the categorized and systematic approach suggested here will provide a proper insight for analyzing risk from a digital system.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4159-4169
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• 2022

Hydrogen mitigation using Passive Autocatalytic Recombiners (PARs) has been widely accepted methodology inside reactor containment of accident struck Nuclear Power Plants. They reduce hydrogen concentration inside reactor containment by recombining it with oxygen from containment air on catalyst surfaces at ambient temperatures. Exothermic heat of reaction drives the product steam upwards, establishing natural convection around PAR, thus invoking homogenisation inside containment. CFD models resolving individual catalyst plate channels of PAR provide good insight about temperature and hydrogen recombination. But very thin catalyst plates compared to large dimensions of the enclosures involved result in intensive calculations. Hence, empirical correlations specific to PARs being modelled are often used in integral containment studies. In this work, an experimentally validated CFD model of PAR has been employed for developing an empirical correlation for Indian PAR. For this purpose, detailed parametric study involving different gas mixture variables at PAR inlet has been performed. For each case, respective values of gas mixture variables at recombiner outlet have been tabulated. The obtained data matrix has then been processed using regression analysis to obtain a set of correlations between inlet and outlet variables. The empirical correlation thus developed, can be easily plugged into commercially available CFD software.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.760-768
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• 2017

The transition temperature shift (TTS) of the reactor pressure vessel materials is an important factor that determines the lifetime of a nuclear power plant. The prediction of the TTS at the end of a plant's lifespan is calculated based on the equation of Regulatory Guide 1.99 revision 2 (RG1.99/2) from the US. The fluence factor in the equation was expressed as a power function, and the exponent value was determined by the early surveillance data in the US. Recently, an advanced approach to estimate the TTS was proposed in various countries for nuclear power plants, and Korea is considering the development of a new TTS model. In this study, the TTS trend of the Korean surveillance test results was analyzed using a nonlinear regression model and a mixed-effect model based on the power function. The nonlinear regression model yielded a similar exponent as the power function in the fluence compared with RG1.99/2. The mixed-effect model had a higher value of the exponent and showed superior goodness of fit compared with the nonlinear regression model. Compared with RG1.99/2 and RG1.99/3, the mixed-effect model provided a more accurate prediction of the TTS.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.1111-1112
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.1170-1171
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• 2005

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.617-648
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• 2009

Under the government supported long-term nuclear R&D program, the severe accident research program at KAERI is directed to investigate unresolved severe accident issues such as core debris coolability, steam explosions, and hydrogen combustion both experimentally and numerically. Extensive studies have been performed to evaluate the in-vessel retention of core debris through external reactor vessel cooling concept for APR1400 as a severe accident management strategy. Additionally, an improvement of the insulator design outside the vessel was investigated. To address steam explosions, a series of experiments using a prototypic material was performed in the TROI facility. Major parameters such as material composition and void fraction as well as the relevant physics affecting the energetics of steam explosions were investigated. For hydrogen control in Korean nuclear power plants, evaluation of the hydrogen concentration and the possibility of deflagration-to-detonation transition occurrence in the containment using three-dimensional analysis code, GASFLOW, were performed. Finally, the integrated severe accident analysis code, MIDAS, has been developed for domestication based on MELCOR. The data transfer scheme using pointers was restructured with the modules and the derived-type direct variables using FORTRAN90. New models were implemented to extend the capability of MIDAS.

• Corrosion Science and Technology
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• v.15 no.5
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• pp.245-252
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• 2016

The pipelines and equipments are degraded by flow-accelerated corrosion (FAC), and a large-scale test facility was constructed for simulate the FAC phenomena in secondary coolant environment of PWR type nuclear power plants. Using this facility, FAC test was performed on weld pipe (carbon steel and low alloy steel) at the conditions of high velocity flow (> 10 m/s). Wall thickness was measured by high temperature ultrasonic monitoring systems (four-channel buffer rod type and waveguide type) during test period and room temperature manual ultrasonic method before and after test period. This work deals with the complex effects of flow velocity on the wall thinning in weld pipe and the test results showed that the higher flow velocity induced different increasement of wall thinning rate for the carbon steel and low alloy steel pipe.

• Nuclear Engineering and Technology
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• v.51 no.5
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• pp.1297-1306
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• 2019

Probabilistic safety assessments (PSA) have been used for several decades to visualize the risk level of commercial nuclear power plants (NPPs). Since the role of a human reliability analysis (HRA) is to provide human error probabilities for safety critical tasks to support PSA, PSA quality is strongly affected by HRA quality. Therefore, it is important to understand the underlying limitations or problems of HRA techniques. For this reason, this study conducted a survey among 14 subject matter experts who represent the HRA community of domestic Korean NPPs. As a result, five significant HRA issues were identified: (1) providing a technical basis for the K-HRA (Korean HRA) method, and developing dedicated HRA methods applicable to (2) diverse external events to support Level 1 PSA, (3) digital environments, (4) mobile equipment, and (5) severe accident management guideline tasks to support Level 2 PSA. In addition, an HRA method to support multi-unit PSA was emphasized because it plays an important role in the evaluation of site risk, which is one of the hottest current issues. It is believed that creating such a catalog of prioritized issues will be a good indication of research direction to improve HRA and therefore PSA quality.

• Corrosion Science and Technology
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• v.22 no.3
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• pp.214-219
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• 2023

Considering the case in the United States where most nuclear power plants with an initial design life of 40 years continue to operate until 60 or 80 years after undergoing material soundness evaluation, it is time to plan a more robust long-term operation strategy for nuclear power plants in Korea. There are some reports that SRO/LRO might be formed when Alloy 690 is heat treated for 10,000 hours to 100,000 hours at 360 to 450 ℃. The possibility of LRO formation in Alloy 690 steam generator tubings of Kori nuclear power plant unit 1 (Kori-1) was investigated using existing research papers. The mechanism in which SRO/LRO occurred was also surveyed. Alloy 690 was found to be more likely to cause ordering than Alloy 600 in terms of alloy composition. The ordering could be evaluated through changes in material properties. However, it is difficult to evaluate it from a microstructural point of view. The likelihood of LRO in Alloy 690 of the Kori-1 plant operated at 320 ℃ for 19 years seemed to be low in terms of time and exposure temperature.