• Proceedings of the Korean Geotechical Society Conference
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• 2000.03b
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• pp.542-547
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• 2000

It has become interested in the concept of permeable barriers for the containment and/or destruction of contaminated groundwater. The purpose of these trench-like barriers is to provide in situ capture and possibly destruction of the contaminant while preserving groundwater flow to uncontaminated zones. For instance, a trichloreethylene(TCE) plume may be contained by a permeable in which reactive iron reduces TCE to ethylene and ethane, compounds which can be easily biodegraded. The objective of this research is to examine the feasibility of using zero-valent iron as a clean-up media in permeable reactive barrier system. A series of laboratory column tests are performed. The concentration of influent and effluent water and the rate of clean up are analysed from these test results. The experimental result shows that the majority of the contamination in groundwater is removed in the reactor. And it shows the corresponding increase in the concentration of chloride ions through the reactor. Results from this study indicate that permeable reactive barrier containing admixtures of zero-valent iron and other materials can effectively clean up groundwater contaminated with organic compounds.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.514-519
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• 1997

The MRX is an integral type ship reactor with 100 MWt power, which is designed by Japan Atomic Energy Research Institute. It is characterized by integral type PWR, in-vessel type control roe drive mechanism, water-filled containment vessel and passive decay heat removal system. Marine reactor should have high passive safety. Therefore, in this study, we simulated the loss of flow accident to verify the passive decay heat removal by natural circulation using RETRAN-03 code. auxiliary feed water systems are used for decay heat removal mechanism and results are compared with the loss of flow accident analysis using emergency decay heat removal system by JAERI. Results are very similar to case of EDRS 1 loop operation in JAERI analysis and decay heat is successfully removed by natural circulation.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.61-66
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• 1996

High pressure. high temperature steam/water blowdown test loop has been constructed. The loop simulates a pressurizer. depressurizalion system and In-Containment Refueling Water Storage Tank (IRWST) with full pressure and temperature conditions. and will be used to generate data for development of an optimal sparser as well as for design of safety/automatic depressurization system. In addition. experiments for reactor safety and pressurizer thermal hydraulics are scheduled. In this paper. general description of the Blowdown and Condensation (B&C) Loop will be given together with the test program.

• 한국전산유체공학회:학술대회논문집
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• 1999.05a
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• pp.205-212
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• 1999

Safety Depressurization System of the Korean Next Generation Reactor prevents the Reactor Coolant System from over-pressurization by discharging the coolant with high pressure and temperature into the In-containment Refueling Water Storage Tank(IRWST) during an accident. If temperature in the IRWST rises above the temperature limit of $200\;^{\circ}F$ due to the discharged coolant, an unstable steam condensation may occur and cause large load on the IRWST wall. To investigate whether this condition can be reached or not for the design basis accident, the flow and temperature distributions of water in the IRWST wire calculated by using CFX 4.2 computer code. The results show that the local water temperature does not exceeds the temperature limit within the transient time of 5 seconds.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.19 no.1
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• pp.587-595
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• 2018

Countries that have been banned from building nuclear plants are becoming more tolerant in response to global warming and climate change. Thus, the construction of future nuclear plants will increase, and the competition will also intensify. A nuclear power plant has a long construction period compared with conventional construction projects. In order to gain a competitive advantage in nuclear power plant construction, the construction period must be decreased. Therefore, the purpose of this study is to develop an exterior wall form for a reactor containment building to reduce the construction time by increasing the height of the form. The structural safety, constructability, and economic feasibility were analyzed to assess the applicability of the proposed form. The proposed form was determined to be structurally safe. Furthermore, the construction period was shortened by reducing the duration of the construction units, and the total construction cost and interest were also reduced. Therefore, the proposed form could contribute to reducing the construction period for nuclear power plants.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.37-50
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• 2020

A passive containment cooling system (PCCS) has been developed as advanced safety feature for innovative power reactor (iPOWER). Passive systems are inherently less stable than active systems and the PCCS encountered the flashing-induced instability previously identified. The objective of this study is to develop stability maps for flashing-induced instability using MARS (Multi-dimensional Analysis of Reactor Safety) code. Firstly, we conducted a series of sensitivity analysis to see the effects of time step size, nodalization, and alternative MARS user options on the onset of flashing-induced instability. The riser nodalization strongly affects the prediction of flashing in a long riser of the PCCS, while time step size and alternative user options do not. Based on the sensitivity analysis, a standard input and an analysis methodology were set up to develop the stability maps of PCCS. We found out that the calculated equilibrium quality at the exit of the riser as a stability boundary above 5 kW/㎡ was approximately 1.2%, which was in good agreement with Furuya's results. However, in case of a very low heat flux condition, the onset of instability occurred at the lower equilibrium quality. In addition, it was confirmed that inlet throttling reduces the unstable region.

• Journal of the Earthquake Engineering Society of Korea
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• v.23 no.3
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• pp.149-157
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• 2019

The purpose of this study is to investigate the effects of the significant duration of ground motions on responses of base-isolated nuclear power plants (NPPs). Two sets of ground motion records with short duration (SD) and long duration (LD) motions, scaled to match the target response spectrum, are used to perform time-history analyses. The reactor containment building in the Advanced Power Reactor 1400 (APR1400) NPP is numerically modeled using lumped-mass stick elements in SAP2000. Seismic responses of the base-isolated NPP are monitored in forms of lateral displacements, shear forces, floor response spectra of the containment building, and hysteretic energy of the lead rubber bearing (LRB). Fragility curves for different limit states, which are defined based on the shear deformation of the base isolator, are developed. The numerical results reveal that the average seismic responses of base-isolated NPP under SD and LD motion sets were shown to be mostly identical. For PGA larger than 0.4g, the mean deformation of LRB for LD motions was bigger than that for SD ones due to a higher hysteretic energy of LRB produced in LD shakings. Under LD motions, median parameters of fragility functions for three limit states were reduced by 12% to 15% compared to that due to SD motions. This clearly indicates that it is important to select ground motions with both SD and LD proportionally in the seismic evaluation of NPP structures.

• Nuclear Engineering and Technology
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• v.54 no.6
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• pp.2077-2083
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• 2022

In the unlikely event of core disruptive accident in sodium cooled fast reactors, the reactor containment building would be bottled up with sodium and fission product aerosols. The behavior of these aerosols is crucial to estimate the in-containment source term as a part of nuclear reactor safety analysis. In this work, the evolution of sodium aerosol characteristics (mass concentration and size) is simulated using HAARM-S code. The code is based on the method of moments to solve the integro-differential equation. The code is updated to FORTRAN-77 and run in Microsoft FORTRAN PowerStation 4.0 (on Desktop). The sodium aerosol characteristics simulated by HAARM-S code are compared with the measured values at Aerosol Test Facility. The maximum deviation between measured and simulated mass concentrations is 30% at initial period (up to 60 min) and around 50% in the later period. In addition, the influence of humidity on aerosol size growth for two different aerosol mass concentrations is studied. The measured and simulated growth factors of aerosol size (ratio of saturated size to initial size) are found to be matched at reasonable extent. Since sodium is highly reactive with atmospheric constituents, the aerosol growth factor depends on the hygroscopic growth, chemical transformation and density variations besides coagulation. Further, there is a scope for the improvement of the code to estimate the aerosol dynamics in confined environment.

• Nuclear Engineering and Technology
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• v.56 no.1
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• pp.132-140
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• 2024

In the event of a severe accident in Sodium Cooled Fast Reactors (SFR), the sodium combustion aerosols along with fission product aerosols would migrate to the environment through leak paths of the Reactor Containment Building (RCB) concrete wall under positive pressure. Understanding the characteristics of sodium aerosol transport through concrete leak paths is important as it governs the environmental source term. In this context, experiments are conducted to study the influence of various parameters like pressure, initial mass concentration, leak path diameter, humidity etc., on the transport and deposition of sodium aerosols in straight leak paths of concrete. The leak paths in concrete specimens are prepared by casting and the diameter of the leak path is measured using thermography technique. Aerosol transport experiments are conducted to measure the transported and plugged aerosol mass in the leak paths and corresponding plugging times. The values of differential pressure, aerosol concentration and relative humidity taken for the study are in the ranges 10-15 kPa, 0.65-3.04 g/m³ and 30-90% respectively. These observations are numerically simulated using 1-Dimensional transport equation. The simulated values are compared with the experimental results and reasonable agreement among them is observed. From the safety assessment view of reactor, the approach presented here is conservative as it is with straight leak paths.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2534-2544
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• 2024

This paper proposes a flooding safety system (FSS) and its operation strategy that can provide long-term safety and effective maintenance for modules of small modular reactor (SMR) and metal containment maintained at dried environment during normal operation. During hypothesized accidents, the FSS re-collects the evaporated steam into the common pool by the condenser installed above the common water pool and provides an emergency coolant for the cavities and auxiliary pools. This study suggested that the condensate re-collection strategy using the FSS can effectively delay the depletion of available water in response to the accidents. Without recollection, the achievable grace periods ranged from 44 to 1507 days for six-module and one-module accidents, respectively. However, with a full re-collection (ratio = 1.0), the time to total depletion of emergency coolant was estimated indefinite. Even with a partial re-collection ratio of 0.3, a grace period of 83.5 days could be ensured for a six-module transient. This study reported the effectiveness of condensate re-collection and the FSS as an innovative safety management strategy and system. Employing a condensate re-collection strategy with a high re-collection ratio can enhance the long-term safety and effective convenience of SMR operations and maintenance.