• Nuclear Engineering and Technology
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• v.18 no.3
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• pp.183-193
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• 1986

Full scale HDR containment experiment series pointed out that the previous containment analysis models have a number of shortcomings. One of them is on the calculational model of short term (0~2sec) pressure difference. The pressure differences between subcompartments are dependent on the flow rate, fluid density, head loss coefficient, and flow area ratio. It, however, is not known that any of them is largely attributed to the disagreement of pressure difference between the measured and the calculated values. In this study, the head loss coefficients are expressed with another form to improve the analytic model. The pressure and the pressure difference are evaluated by using COMPARE code with new correlation, and the results show better agreements with experimental values for V.42 test, but overestimate the measured values for V, 43 and underestimate for V.44.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.183-190
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• 1996

Leak-before-break(LBB) approach has been shown to be both cost and risk effective by reducing maintenance cost and occupational exposure when applied to high energy piping in nuclear power plants. For Korean Next Generation Reactor(KNGR) development, LBB is considered for the Main Steam Line(MSL) piping inside containment. Unlike the reactor coolant piping leakages which can be detected by particulate and gaseous radiation monitoring, main steam line leak detection systems must be based on principles that do not involve radioactivity. Ceramics are widely used as humidity sensor materials which can be further developed for nuclear applications. In this paper, we describe the progress in the development of ceramic humidity sensors for use with the main steam lines of KNGR.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2511-2521
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• 2020

The Fukushima accident is characterized by the fact that three reactors at the same site experienced reactor vessel failure and the accident resulted in significant radiological release to the environment, which was about 1/10 of the Chernobyl releases. The safe removal of fuel debris in the reactor vessel and Primary Containment Vessel (PCV) and treatment of huge amount of contaminated water are the major issues for the decommissioning in coming decades. Discussions on the new researches efforts being carried out in the area of investigation of the end state of fuel debris and Boling Water reactor (BWR) specific core melt progression, development of technologies for the mitigation of radiological releases to comply with the strengthened safety requirement set after the Fukushima accident are discussed.

• Journal of the Korean Society of Visualization
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• v.21 no.2
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• pp.45-54
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• 2023

In the present study, we propose new type of a spinning disk reactor(SDR) with high performance and very convenient structure to make a large scale equipment from lab-scale than the conventional one. A split-disk experimental equipment, based on new type of spinning disk reactor, has been developed to generate an energy to break a bulk of injected gas into smaller gas bubble. Several cases of an experimental observation make it to confirm that a bulk of injecting gas could be continuously break into smaller bubbles. It shows the feasibility to make a scale-up of SDR by using the characteristic of Taylor-Proudman column in rotating flow. A theoretical study on single phase liquid flow is given to predict a liquid induced shear stress, which make the present study to be self-containment.

• Nuclear Engineering and Technology
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• v.50 no.5
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• pp.665-672
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• 2018

A venturi scrubber is an important element of Filtered Containment Venting System (FCVS) for the removal of aerosols in contaminated air. The present work involves computational fluid dynamics (CFD) study of dust particle removal efficiency of a venturi scrubber operating in self-priming mode using ANSYS CFX. Titanium oxide ($TiO_2$) particles having sizes of 1 micron have been taken as dust particles. CFD methodology to simulate the venturi scrubber has been first developed. The cascade atomization and breakup (CAB) model has been used to predict deformation of water droplets, whereas the Eulerian-Lagrangian approach has been used to handle multiphase flow involving air, dust, and water. The developed methodology has been applied to simulate venturi scrubber geometry taken from the literature. Dust particle removal efficiency has been calculated for forced feed operation of venturi scrubber and found to be in good agreement with the results available in the literature. In the second part, venturi scrubber along with a tank has been modeled in CFX, and transient simulations have been performed to study self-priming phenomenon. Self-priming has been observed by plotting the velocity vector fields of water. Suction of water in the venturi scrubber occurred due to the difference between static pressure in the venturi scrubber and the hydrostatic pressure of water inside the tank. Dust particle removal efficiency has been calculated for inlet air velocities of 1 m/s and 3 m/s. It has been observed that removal efficiency is higher in case of higher inlet air velocity.

• Journal of Energy Engineering
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• v.24 no.3
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• pp.96-108
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• 2015

A passive containment cooling system has been designed to remove the heat inside a containment during accidents without external power supply. In this work, the PCCS was introduced in the APR1400 plant to replace the containment spray system and, then, the thermal-hydraulic performance of the PCCS was analyzed using the system thermal-hydraulic computer code, MARS. A double-ended cold-leg break accident, which is known to induce the maximum pressure in the containment, is simulated, where the thermal hydraulics of the PCCS, the reactor coolant system, and the containment are simultaneously simulated. The results of the calculations showed that the PCCS can replace the existing spray system and that the containment building and its internal structure also play a very important role for the heat removal during the accident. Some sensitivity calculations were carried out to evaluate the model uncertainty and the effects of design parameters. The limitations of the PCCS are also discussed.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.34 no.3
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• pp.167-174
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• 2021

Recent investigation into the integrity of nuclear containment buildings has highlighted the importance of developing an elaborate diagnostic method to evaluate the distribution and size of cavities inside concrete walls. As part of developing such a method, this paper presents a finite element approach to modeling elastic waves propagating in the containment building walls of a nuclear power plant. We introduce a perfectly matched layer (PML) wave-absorbing boundary to limit the large-scale nuclear containment wall to the region of interest. The formulation results in a semi-discrete form with symmetric damping and stiffness matrices. The transient elastic wave equations for a mixed unsplit-field PML were solved for displacement and stresses in the time domain. Numerical results show that the sensitivity of displacement, velocity, acceleration, and stresses is large depending on the size and location of the cavity. The dynamic response of the wall slightly differs depending on the existence of the containment liner plate. The results of this study can be applied to a full-waveform inversion approach for characterizing cavities inside a containment wall.

• Journal of the Earthquake Engineering Society of Korea
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• v.25 no.2
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• pp.83-91
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• 2021

Seismic designs for Korean nuclear power plants (NPPs) under earthquakes' design basis are noticed due to the recent earthquake events in Korea and Japan. Japan has developed the technologies and experiences of the NPPs through theoretical research and experimental verification with extensively accumulated measurement data. This paper describes the main features of the design-time history complying with the Japanese seismic design standard. Proper seed motions in the earthquake catalog are used to generate one set of design time histories. A magnitude and epicentral distance specify the amplitude envelope function configuring the shape of the earthquake. Cumulative velocity response spectral values of the design time histories are compared and checked to the target response spectra. Spectral accelerations of the time histories and the multiple-damping target response spectra are also checked to exceed. The generated design time histories are input to the reactor building seismic analyses with fixed-base boundary conditions to calculate the seismic responses. Another set of design time histories is generated to comply with Korean seismic design procedures for NPPs and used for seismic input motions to the same reactor containment building seismic analyses. The responses at the dome apex of the building are compared and analyzed. The generated design time histories will be also applied to subsequent seismic analyses of other Korean standard NPP structures.

• Proceedings of the Korea Concrete Institute Conference
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• 2002.05a
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• pp.475-480
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• 2002

Tension tests of six half-thickness concrete containment wall elements were conducted as part of a Korea Atomic Energy Research Institute (KARRI) program. The aim of the KAERI test program is providing a test-verified analytical method for estimating capacities of concrete reactor containment buildings under internal overpressurization from postulated degraded core accidents. The data from the tests reported herein should be useful for benchmarking method that requires modeling of material behavior including concrete cracking and reinforcement/concrete interaction exhibited by the test. Major test variable is the compressive strength of concrete and its effect on the behavior of prestressed concrete panel subjected to biaxial tension.

• Journal of Advanced Marine Engineering and Technology
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• v.40 no.2
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• pp.112-119
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• 2016

Many studies have been conducted on the performance of a passive autocatalytic recombiner (PAR), but not many have focused on the locations where the PAR is installed. During a severe accident in a nuclear reactor containment, a large amount of hydrogen gas can be produced and released into the containment, leading to hydrogen deflagration or a detonation. A PAR is a hydrogen mitigation method that is widely implemented in current and advanced light water reactors. Therefore, for this study, a PAR was installed at different locations in order to investigate the difference in hydrogen reduction rate. The results indicate that the hydrogen reduction rate of a PAR is proportional to the distance between the hydrogen induction location and the bottom wall.