• Journal of Soil and Groundwater Environment
• /
• v.16 no.1
• /
• pp.62-70
• /
• 2011

This study was conducted to evaluate the performance of soil-cement walls (SCWs) to control leachate from a leaking landfill site. Tracer tests revealed that the SCW was effective to control groundwater seepage. Approximately two-months of curing period appeared to be sufficient to ensure thorough containment of landfill leachate, although a three-week period was not enough. The water quality of the monitoring wells after construction of the SCWs met the groundwater quality standard of the korean Waste Management Act, except for bacteria and coliform groups. Also an analysis of a spring water around the landfill showed that the concentrations of ammonia, inorganic nitrogen and soluble manganese which had been common contaminants in the spring water decreased dramatically after constructing the walls. Therefore, the results suggested that a SCW can be an attractive method to control leachate from a leaking landfill site.

• Nuclear Engineering and Technology
• /
• v.42 no.1
• /
• pp.1-8
• /
• 2010

Geological disposal is designed to provide safe containment of radioactive waste for very long times, with the containment provided by a combination of engineered and geological barriers. In the extreme long term, after many thousands of years or longer, residual amounts of long-lived radionulides such as Cl-36, but also radionuclides in the natural decay chains, may be released into the environment normally accessed and used by humans, termed here, the biosphere. It is necessary to ensure that any such releases meet radiation protection objectives through the development of a safety case, which will include assessment of radiation doses to humans. The design of such dose calculations over such long timeframes is not straightforward, because of the range of potentially relevant assumptions which could be made, concerning environmental change and changes in human behavior. These conceptual uncertainties are additional to those that more typically arise, for example, in the assessment of present day situations, but which also have to be addressed. The issue has therefore been subject to international cooperation for many years. This paper summarizes the evolution and results of that collaboration leading up to the present day, taking account of developments in international recommendations on radiation protection objectives and the more recent greater focus on preparation of site specific safety cases.

• Nuclear Engineering and Technology
• /
• v.54 no.1
• /
• pp.386-400
• /
• 2022

Seismic fragility analysis, a part of seismic probabilistic risk assessment (SPRA), is commonly used to establish the relationship between a representative property of earthquakes and the failure probability of a structure, component, or system. Current guidelines on the SPRA of nuclear power plants (NPPs) used worldwide mainly reflect the earthquake characteristics of the western United States. However, different earthquake characteristics may have a significant impact on the seismic fragility of a structure. Given the concern, this study aimed to investigate the effects of earthquake characteristics on the seismic fragility of concrete containments housing the OPR-1000 reactor. Earthquake time histories were created from 30 ground motions (including those of the 2016 Gyeongju earthquake) by spectral matching to the site-specific response spectrum of Hanbit nuclear power plants in South Korea. Fragility curves of the containment structure were determined under the linear response history analysis using a lumped-mass stick model and 30 ground motions, and were compared in terms of earthquake characteristics. The results showed that the median capacity and high confidence of low probability of failure (HCLPF) tended to highly depend on the sustained maximum acceleration (SMA), and increase when using the time histories which have lower SMA compared with the others.

• Nuclear Engineering and Technology
• /
• v.54 no.6
• /
• pp.2297-2310
• /
• 2022

A condensation heat transfer model is essential to accurately predict the performance of the passive containment cooling system (PCCS) during an accident in an advanced light water reactor. However, most of existing models tend to predict condensation heat transfer very well for a specific range of thermal-hydraulic conditions. In this study, a new correlation for condensation heat transfer coefficient (HTC) is presented using machine learning technique. To secure sufficient training data, a large number of pseudo data were produced by using ten existing condensation models. Then, a neural network model was developed, consisting of a fully connected layer and a convolutional neural network (CNN) algorithm, DenseNet. Based on the hold-out cross-validation, the neural network was trained and validated against the pseudo data. Thereafter, it was evaluated using the experimental data, which were not used for training. The machine learning model predicted better results than the existing models. It was also confirmed through a parametric study that the machine learning model presents continuous and physical HTCs for various thermal-hydraulic conditions. By reflecting the effects of individual variables obtained from the parametric analysis, a new correlation was proposed. It yielded better results for almost all experimental conditions than the ten existing models.

• Nuclear Engineering and Technology
• /
• v.55 no.9
• /
• pp.3194-3201
• /
• 2023

In water-cooled power reactor, hydrogen is generated in case of steam zirconium reaction during severe accident condition and later on in addition to hydrogen; CO is also generated during molten corium concrete interaction after reactor pressure vessel failure. Passive Autocatalytic Recombiners (PARs) are provided in the containment for hydrogen management. The performance of the PARs in presence of hydrogen and carbon monoxide along with air has been evaluated. Depending on the conditions, CO may either react with oxygen to form carbon dioxide (CO2) or act as catalyst poison, reducing the catalyst activity and hence the hydrogen conversion efficiency. CFD analysis has been carried out to determine the effect of CO on catalyst plate temperature for 2 & 4% v/v H₂ and 1-4% v/v CO with air at the recombiner inlet for a reported experiment. The results of CFD simulations have been compared with the reported experimental data for the model validation. The reaction at the recombiner plate is modelled based on diffusion theory. The developed CFD model has been used to predict the maximum catalyst temperature and outlet species concentration for different inlet velocity and temperatures of the mixture gas. The obtained results were used to fit a correlation for obtaining removal rate of carbon monoxide inside PAR as a function of inlet velocity and concentrations.

• 한국정보디스플레이학회:학술대회논문집
• /
• 2007.08a
• /
• pp.85-86
• /
• 2007

As the manufacturing capacity needs for large size LCD TV shifts very fast into next generation, equipment manufactures face more challenges in development of the systems which can accommodate productivity, reliability and high process quality requirements from the panel makers. AKT committed to continue its contribution to the growth of the LCD market by providing excellent PECVD products. The cost containment and performance improvement are key challenges for large size processing equipment development and this presentation discusses various challenges and AKT's solutions in developing large size PECVD equipment beyond Gen. 8.

• Nuclear Engineering and Technology
• /
• v.56 no.3
• /
• pp.803-811
• /
• 2024

Nuclear accidents such as Fukushima Daiichi have highlighted the potential of passive safety systems to replace or complement active safety systems as part of the overall prevention and/or mitigation strategies. In addition, passive systems are key features of Small Modular Reactors (SMRs), for which they are becoming almost unavoidable and are part of the basic design of many reactors available in today's nuclear market. Nevertheless, their potential to significantly increase the safety of nuclear power plants still needs to be strengthened, in particular the ability of computer codes to determine their performance and reliability in industrial applications and support the safety demonstration. The PASTELS project (September 2020-February 2024), funded by the European Commission "Euratom H2020" programme, is devoted to the study of passive systems relying on natural circulation. The project focuses on two types, namely the SAfety COndenser (SACO) for the evacuation of the core residual power and the Containment Wall Condenser (CWC) for the reduction of heat and pressure in the containment vessel in case of accident. A specific design for each of these systems is being investigated in the project. Firstly, a straight vertical pool type of SACO has been implemented on the Framatome's PKL loop at Erlangen. It represents a tube bundle type heat exchanger that transfers heat from the secondary circuit to the water pool in which it is immersed by condensing the vapour generated in the steam generator. Secondly, the project relies on the CWC installed on the PASI test loop at LUT University in Finland. This facility reproduces the thermal-hydraulic behaviour of a Passive Containment Cooling System (PCCS) mainly composed of a CWC, a heat exchanger in the containment vessel connected to a water tank at atmospheric pressure outside the vessel which represents the ultimate heat sink. Several activities are carried out within the framework of the project. Different tests are conducted on these integral test facilities to produce new and relevant experimental data allowing to better characterize the physical behaviours and the performances of these systems for various thermo-hydraulic conditions. These test programmes are simulated by different codes acting at different scales, mainly system and CFD codes. New "system/CFD" coupling approaches are also considered to evaluate their potential to benefit both from the accuracy of CFD in regions where local 3D effects are dominant and system codes whose computational speed, robustness and general level of physical validation are particularly appreciated in industrial studies. In parallel, the project includes the study of single and two-phase natural circulation loops through a bibliographical study and the simulations of the PERSEO and HERO-2 experimental facilities. After a synthetic presentation of the project and its objectives, this article provides the reader with findings related to the physical analysis of the test results obtained on the PKL and PASI installations as well an overall evaluation of the capability of the different numerical tools to simulate passive systems.

• Journal of the Society of Naval Architects of Korea
• /
• v.60 no.4
• /
• pp.222-230
• /
• 2023

In the present work, the scale effect on the Boil-Off Rate (BOR) was investigated based on an analytical method to systematically evaluate the thermal performance of a Liquefied Natural Gas (LNG) Cargo Containment System (CCS). A two-dimensional thermal resistance network model was developed to accurately estimate the heat ingress into the CCS from the outside. The analysis was performed for the KC-1 LNG membrane tank under the IGC and USCG design conditions. The ballast compartment of both the LNG tank and cofferdam was divided into six sections and a thermal resistance network model was made for each section. To check the validity of the developed model, the analysis results were compared with those from existing literature. It was shown that the BOR values under the IGC and USCG design conditions were agreed well with previous numerical results with a maximum error of 1.03% and 0.60%, respectively. A SDR, the scale factor of the LNG CCS was introduced and the BOR, air temperature of the ballast compartment, and the surface temperature of the inner hull were obtained to examine the influence of the SDR on the thermal performance. Finally, a correlation for the BOR was proposed, which could be expressed as a simple formula inversely proportional to the SDR. The proposed correlation could be utilized for predicting the BOR of a full-scale LNG tank based on the BOR measurement data of lab-scale model tanks.

• KSCE Journal of Civil and Environmental Engineering Research
• /
• v.13 no.2
• /
• pp.21-29
• /
• 1993

In this paper, a probability-based reliability analysis was proposed based on a finite element method-based random vibration analysis and serviceability limit state of structures. The limit state model defined for the study is a serviceability limit state in terms of the more realistic crack failure that might cause the emission of radioactive materials. The SAP V-2 is used for a three-dimensional finite element analysis of concrete containment structure, and the reliability analysis is carried out by modifying HRAS reliability analysis program for this study. In this study, the load factors for the design of reinforced concrete cointainment structures in Korea are proposed by considering appropriate load combination criteria for design, and the results are compared with the present ASME code. The proposed load factors were proved to be in accordance with a set of code performance objective and showed consistency in the limit state probability.

• Nuclear Engineering and Technology
• /
• v.47 no.1
• /
• pp.33-46
• /
• 2015

During the past 10 years, the Korea Atomic Energy Research Institute (KAERI) has performed a study to control hydrogen gas in the containment of the nuclear power plants. Before the Fukushima accident, analytical activities for gas distribution analysis in experiments and plants were primarily conducted using a multidimensional code: the GASFLOW. After the Fukushima accident, the COM3D code, which can simulate a multidimensional hydrogen explosion, was introduced in 2013 to complete the multidimensional hydrogen analysis system. The code validation efforts of the multidimensional codes of the GASFLOW and the COM3D have continued to increase confidence in the use of codes using several international experimental data. The OpenFOAM has been preliminarily evaluated for APR1400 containment, based on experience from coded validation and the analysis of hydrogen distribution and explosion using the multidimensional codes, the GASFLOW and the COM3D. Hydrogen safety in nuclear power has become a much more important issue after the Fukushima event in which hydrogen explosions occurred. The KAERI is preparing a large-scale test that can be used to validate the performance of domestic passive autocatalytic recombiners (PARs) and can provide data for the validation of the severe accident code being developed in Korea.