• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.372.2-372
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.182.2-182
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.795-796
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• 2005

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1164-1171
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• 2020

T-junction structures play an important role in nuclear power plant systems. Research on liquid entrainment is mostly based on small-scale branch pipes (d/D ≤ 0.2) and attention paid to large-scale branch pipes (0.33 < d/D < 1) is insufficient. Accordingly, this study implements a series of experiments on the liquid entrainment of T-junction with different angles (32.2°,47.9°,62.3°,90°) through a large-scale branch (d/D = 0.675). The onset liquid entrainment is related to the gas phase Froude number Fr_g, the dimensionless gas chamber height h_b/d and the branch pipe angle 𝜃. As Fr_g increases, h_b/d also rises. With a constant h_b/d, the onset liquid entrainment changes from droplets entrainment by the gas phase to that by the rising liquid film. The steady-state liquid entrainment is related to w_3g, h/d and 𝜃. With constant w_3g and h/d, the branch quality grows as the branch angle increases. With a certain h/d, the branch quality increases, as the w_3g number increases.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.1
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• pp.51-58
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• 2006

Since 2002, more than 400 PWR spent nuclear fuel assemblies have been transported and stored on-site using transport casks in order to secure the storage capacity of PWR spent nuclear fuel of Kori nuclear power plant. The complete on-site transport system, which includes KN-12 transport casks, the related equipment and transport vehicles, had been developed and provided. KN-12 transport casks were designed, fabricated and licensed in accordance with Korean and IAEA's transport regulations, and the related equipment was also provided in accordance with the related regulations. The on-site transport and storage operation using two KN-12 casks and the related equipment has been conducted, and the strict Quality Control and Radiation Safety Management through the whole process has been carried out so as to achieve the required safety and reliability of the on-site transport of spent nuclear fuel.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1577-1586
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• 2023

Safety qualification of a steam generator is a crucial issue related to faulted condition design loads, including earthquake loads, and it should be ensured that the structural integrity of a steam generator does not exceed its design load. Using data from the Gyeongju and Pohang earthquakes, the two most powerful recorded seismic events in Korea, seismic analyses of a typical steam generator are conducted in this study. The modal characteristics are used to develop an input deck for these analyses. With a time history analysis, the responses of the steam generator in the event of an earthquake are obtained. In particular, the displacement, velocity, and acceleration responses are obtained in the time domain, with these outcomes then used for a detailed structural analysis as part of the ensuing assessment. The response spectra are also generated to determine the response characteristics in the frequency domain, focusing on the response comparisons between the Gyeongju and Pohang earthquakes. Structural integrity can be ensured by performing additional analysis using results obtained from the time history analysis considering the input excitations of various earthquakes considered in the design.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.777-782
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• 1998

To terminate the progression of the sever accident and mitigate the accident consequences, corium coaling has been suggested as one of most important design features considered in the swore accident mitigation. Till now, some kinds of cooling methodologies have been identified and, specially the corium cooling at the reactor cavity has been considered as one of the most promising cooling methodologies. Moreover, several design requirements related to the cerium cooling at the reactor cavity have been also suggested and applied to the design of the next generation reactor. In this study technical description are briefly described for the important issues related to the cerium cooling at the reactor cavity, i.e. cavity area, cavity flooding system, etc., and simple evaluation for those items have been performed considering present technical levels the experiment and analytical works..

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1032-1036
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• 2019

When the equipment which is related to safety or important to power production is installed in nuclear power plant units (NPPs), verification of equipment Electromagnetic Susceptibility (EMS) must be performed. The low-frequency radiated magnetic field susceptibility (RS101) test is one of the EMS tests specified in U.S NRC (Nuclear Regulatory Commission) Regulatory Guide (RG) 1.180 revision 1. The RS101 test verifies the ability of equipment installed in close proximity to sources of large radiated magnetic fields to withstand them. However, RG 1.180 revision 1 allows for an exemption of the low-frequency radiated magnetic susceptibility (RS101) test if the safety-related equipment will not be installed in areas with strong sources of magnetic fields. There is no specific exemption criterion in RG 1.180 revision 1. EPRI TR-102323 revision 4 specifically provides a guide that the low-frequency radiated magnetic field susceptibility (RS101) test can be conservatively exempted for equipment installed at least 1 m away from the sources of large magnetic fields (>300 A/m). But there is no exemption criterion for equipment installed within 1 m of the sources of smaller magnetic fields (<300 A/m). Since some types of equipment radiating magnetic flux are often installed near safety related equipment in an electrical equipment room (EER) and main control room (MCR), the RS101 test exemption criterion needs to be reasonably defined for the cases of installation within 1 m. There is also insufficient data regarding the strength of magnetic fields that can be used in NPPs. In order to ensure confidence in the RS101 test exemption criterion, we measured the strength of low-frequency radiated magnetic fields by distance. This study is expected to provide an insight into the RS101 test exemption criterion that meets the RG 1.180 revision 1. It also provides a margin analysis that can be used to mitigate the influence of low-frequency radiated magnetic field sources in NPPs.

• Nuclear Engineering and Technology
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• v.46 no.5
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• pp.569-580
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• 2014

Seismic isolation is a viable strategy for protecting safety-related nuclear structures from the effects of moderate to severe earthquake shaking. Although seismic isolation has been deployed in nuclear structures in France and South Africa, it has not seen widespread use because of limited new build nuclear construction in the past 30 years and a lack of guidelines, codes and standards for the analysis, design and construction of isolation systems specific to nuclear structures. The funding by the United States Nuclear Regulatory Commission of a research project to the Lawrence Berkeley National Laboratory and MCEER/University at Buffalo facilitated the writing of a soon-to-be-published NUREG on seismic isolation. Funding of MCEER by the National Science Foundation led to research products that provide the technical basis for a new section in ASCE Standard 4 on the seismic isolation of safety-related nuclear facilities. The performance expectations identified in the NUREG and ASCE 4 for seismic isolation systems, and superstructures and substructures are described in the paper. Robust numerical models capable of capturing isolator behaviors under extreme loadings, which have been verified and validated following ASME protocols, and implemented in the open source code OpenSees, are introduced.

• Nuclear Engineering and Technology
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• v.41 no.10
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• pp.1361-1374
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• 2009

Due to the need of data for a Human Reliability Analysis (HRA), a number of data collection efforts have been undertaken in several different organizations. As a part of this effort, a human error analysis that focused on a set of simulator records on a Steam Generator Tube Rupture (SGTR) scenario was performed by using the Human Event Repository and Analysis (HERA) system. This paper summarizes the process and results of the HERA analysis, including discussions about the usability of the HERA system for a human error analysis of simulator data. Five simulated records of an SGTR scenario were analyzed with the HERA analysis process in order to scrutinize the causes and mechanisms of the human related events. From this study, the authors confirmed that the HERA was a serviceable system that can analyze human performance qualitatively from simulator data. It was possible to identify the human related events in the simulator data that affected the system safety not only negatively but also positively. It was also possible to scrutinize the Performance Shaping Factors (PSFs) and the relevant contributory factors with regard to each identified human event.