• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.805-819
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• 2016

The numerical simulation methodologies to evaluate the structural behaviors of prestressed concrete containment vessels (PCCVs) have been substantially developed in recent decades. However, there remain several issues to be investigated more closely to narrow the gap between test results and numerical simulations. As one of those issues, the effects of no stiffness inside unbonded tendon ducts on the behavior of PCCVs are investigated in this study. Duct holes for prestressing cables' passing are provided inside the containment wall and dome in one to three directions for general PCCVs. The specific stress distribution along the periphery of the prestressing duct hole and the loss of stiffness inside the hole, especially in an unbonded tendon system, are usually neglected in the analysis of PCCVs with the assumption that the duct hole is filled with concrete. However, duct holes are not small enough to be neglected. In this study, the effects of no stiffness inside the unbonded tendon system on the behaviors of PCCVs are evaluated using both analytical and numerical approaches. From the results, the effects of no stiffness in unbonded tendons need to be considered in numerical simulations for PCCVs, especially under internal pressure loading.

• Nuclear Engineering and Technology
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• v.44 no.8
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• pp.871-888
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• 2012

This paper aims at presenting the state of the art, the recent progress, and the perspective for the future, in the modelling of two-phase flow in the horizontal legs of a PWR. All phenomena relevant for safety analysis are listed first. The selection of the modelling approach for system codes is then discussed, including the number of fluids or fields, the space and time resolution, and the use of flow regime maps. The classical two-fluid six-equation one-pressure model as it is implemented in the CATHARE code is then presented and its properties are described. It is shown that the axial effects of gravity forces may be correctly taken into account even in the case of change of the cross section area or of the pipe orientation. It is also shown that it can predict both fluvial and torrential flow with a possible hydraulic jump. Since phase stratification plays a dominant role, the Kelvin-Helmholtz instability and the stability of bubbly flow regime are discussed. A transition criterion based on a stability analysis of shallow water waves may be used to predict the Kelvin-Helmholtz instability. Recent experimental data obtained in the METERO test facility are analysed to model the transition from a bubbly to stratified flow regime. Finally, perspectives for further improvement of the modelling are drawn including dynamic modelling of turbulence and interfacial area and multi-field models.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.781-790
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• 2012

The HyperGam program was developed for the analysis of complex HPGe ${\gamma}$-ray spectra. The previous version of HyperGam was mainly limited to the analysis of ${\gamma}$-ray peaks and the manual logging of the result. In this study, it is specifically developed into a tool for the isotopic analysis of spectra. The newly developed features include nuclide identification and activity determination. An algorithm for nuclide identification was developed to identify the peaks in the spectrum by considering the yield, efficiency, energy and peak area for the ${\gamma}$-ray lines emitted from the radionuclide. The detailed performance of nuclide identification and activity determination was accessed using the IAEA 2002 set of test spectra. By analyzing the test spectra, the numbers of radionuclides identified truly (true hit), falsely (false hit) or missed (misses) were counted and compared with the results from the IAEA 2002 tests. The determined activities of the radionuclides were also compared for four test spectra of several samples. The result of the performance test is promising in comparison with those of the well-known software packages for ${\gamma}$-ray spectrum analysis.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.817-824
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• 2012

The most important component of a nuclear power plant is its nuclear reactor. Studies on the integrity of reactors have become an important part regarding the safety of a nuclear power plant. The US Nuclear Regulatory Commission Regulatory Guide (NRC RG) 1.20 presents a Comprehensive Vibration Assessment Program (CVAP) to be used to verify the structural integrity of the Reactor Vessel Internals (RVI) for flow-induced vibration prior to commercial operation. However, there are few published studies related to the RVI CVAP. We classified the Advanced Power Reactor 1400 (APR1400) RVI CVAP as a non-prototype category-2 reactor as part of an independent validation of its design. The aim of this paper is to present the results of structural response analyses of the Upper Guide Structure (UGS) assembly of the APR1400 reactor. These results show that the UGS and the Inner Barrel Assembly (IBA) meet the specified integrity levels of the design acceptance criteria. The vibration and stress analysis results in this paper will be used as basic information to select measurement locations of the vibration and stress for the APR1400 RVI CVAP.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.745-752
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• 2013

For thermo-fluid and safety analyses of a High Temperature Gas-cooled Reactor (HTGR), intensive efforts are in progress in the developments of the GAMMA+ code of Korea Atomic Energy Research Institute (KAERI) and the AGREE code of the University of Michigan (U of M). One of the important requirements for GAMMA+ and AGREE is an accurate modeling capability of a bypass flow in a prismatic core. Recently, a series of air experiments were performed at Seoul National University (SNU) in order to understand bypass flow behavior and generate an experimental database for the validation of computer codes. The main objective of the present work is to validate the GAMMA+ and AGREE codes using the experimental data published by SNU. The numerical results of the two codes were compared with the measured data. A good agreement was found between the calculations and the measurement. It was concluded that GAMMA+ and AGREE can reliably simulate the bypass flow behavior in a prismatic core.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.811-820
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• 2013

SiC has been under investigation as a potential cladding for LWR fuel, due to its high melting point and drastically reduced chemical reactivity with liquid water, and steam at high temperatures. As SiC is a brittle material its behavior during the reflood phase of a Loss of Coolant Accident (LOCA) is another important aspect of SiC that must be examined as part of the feasibility assessment for its application to LWR fuel rods. In this study, an experimental assessment of thermal shock performance of a monolithic alpha phase SiC tube was conducted by quenching the material from high temperature (up to $1200^{\circ}C$) into room temperature water. Post-quenching assessment was carried out by a Scanning Electron Microscopy (SEM) image analysis to characterize fractures in the material. This paper assesses the effects of pre-existing pores on SiC cladding brittle fracture and crack development/propagation during the reflood phase. Proper extension of these guidelines to an SiC/SiC ceramic matrix composite (CMC) cladding design is discussed.

• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.360-372
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• 2017

Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.

• Journal of Environmental Impact Assessment
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• v.5 no.1
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• pp.107-121
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• 1996

The objective of this study is to establish the application method of environmental information system which is related to hazard identification for Health Risk Assessment. For establishing the environmental information system, fourteen hazardous chemicals were chosen and applicated to the database network such as RTKNET(Right Know Net), MSDS(Material Safety Data Sheets), TRI(Toxic Release Inventory), IRIS, AIRS, etc. The searching method of environmental information is classified to three sections such as the domestic commercial information company, international database agencies, and internet. Recently the importance of environmental information is being emphasized because it is essential 10 use database system in the field of environmental studies. Most of the foreign research organizations are communicating actively for information exchange, and the improvement of the quality of research. It is required to accumulate the data and develop them to database for future research.

• Journal of Families and Better Life
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• v.31 no.2
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• pp.59-75
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• 2013

The purpose of this qualitative study was to explore how divorced mothers had decided to take custody of their children and became single mothers. The experiences of their lives after divorce were also explored. Data were collected from 17 Korean divorced mothers who were divorced between 2004 and 2009, and were raising at least one minor child. The data were analyzed based on the phenomenological data analysis method. Three main themes were identified: (a) reasons for deciding to have physical custody of the children, (b) mothers' experiences of adjustment after divorce, and (c) mothers' need for a policy concerning the well-being of their families. According to the divorced mothers, they decided to have physical custody of the children since they believed raising children was their natural duty of mothers or they were the most appropriate ones to raise the children rather than the fathers. While the mothers were satisfied with their lives after divorce in general, they also experienced difficulties including child care and financial strain. In particular, most mothers experienced work-family conflict related to the lack of reliable child care. When their family lives and work lives collided, the mothers put their children first and chose jobs that helped them take care of their children at the same time. The divorced single mothers hoped that the social safety net for single parents would expand to support their independence. Implications for single-parent policy are discussed.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.267-283
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• 2015

The main objective of this study is to investigate experimentally the two-phase flow characteristics in terms of the direct contact condensation of a steam-water stratified flow in a horizontal rectangular channel. Experiments were performed for both air-water and steam-water flows with a cocurrent flow configuration. This work presents the local temperature and velocity distributions in a water layer as well as the interfacial characteristics of both condensing and noncondensing fluid flows. The gas superficial velocity varied from 1.2 m/s to 2.0 m/s for air and from 1.2 m/s to 2.8 m/s for steam under a fixed inlet water superficial velocity of 0.025 m/s. Some advanced measurement methods have been applied to measure the local characteristics of the water layer thickness, temperature, and velocity fields in a horizontal stratified flow. The instantaneous velocity and temperature fields inside the water layer were measured using laser-induced fluorescence and particle image velocimetry, respectively. In addition, the water layer thickness was measured through an ultrasonic method.