• Journal of the Computational Structural Engineering Institute of Korea
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• v.26 no.2
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• pp.157-164
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• 2013

In this study, the structural performance of nuclear power plant containment buildings, which are made of steel fiber reinforced concrete(SFRC) and subject to aircraft crash, is examined by finite element analyses. The applied loads by aircraft crash against the buildings are modeled using Riera impact load function and by the varying aircraft contact area with respect to time. CSCM concrete model in LS-DYNA is employed to model SFRC. The parameters for the material model are determined from SFRC strength prediction models. Based on the volume ratio of steel fiber in SFRC, the structural performance of nuclear containment buildings subject to aircraft crash are analysed using a commercial finite element analysis program LS-DYNA. The safety assessments of the buildings subject to the crash are discussed and the effectiveness of SFRC for nuclear power plant containment building on the increase of aircraft crash resistance is also evaluated.

• Nuclear Engineering and Technology
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• v.49 no.2
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• pp.426-433
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• 2017

This paper describes the development process, the innovative techniques used and insights gained from the latest integrated, full scope, multistate Level 2 PSA analysis conducted at the Leibstadt Nuclear Power Plant (KKL), Switzerland. KKL is a modern single-unit General Electric Boiling Water Reactor (BWR/6) with Mark III Containment, and a power output of $3600MW_{th}/1200MW_e$, the highest among the five operating reactors in Switzerland. A Level 2 Probabilistic Safety Assessment (PSA) analyses accident phenomena in nuclear power plants, identifies ways in which radioactive releases from plants can occur and estimates release pathways, magnitude and frequency. This paper attempts to give an overview of the advanced modeling techniques that have been developed and implemented for the recent KKL Level 2 PSA update, with the aim of systematizing the analysis and modeling processes, as well as complying with the relatively prescriptive Swiss requirements for PSA. The analysis provides significant insights into the absolute and relative importances of risk contributors and accident prevention and mitigation measures. Thanks to several newly developed techniques and an integrated approach, the KKL Level 2 PSA report exhibits a high degree of reviewability and maintainability, and transparently highlights the most important risk contributors to Large Early Release Frequency (LERF) with respect to initiating events, components, operator actions or seismic component failure probabilities (fragilities).

• Journal of the Earthquake Engineering Society of Korea
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• v.17 no.2
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• pp.53-59
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• 2013

Several researches have been studied to enhance the seismic performance of nuclear power plants (NPPs) by application of seismic isolation. If a seismic base isolation system is applied to NPPs, seismic performance of nuclear power plants should be reevaluated considering the soil-structure interaction effect. The seismic fragility analysis method has been used as a quantitative seismic safety evaluation method for the NPP structures and equipment. In this study, the seismic performance of an isolated NPP is evaluated by seismic fragility curves considering the soil-structure interaction effect. The designed seismic isolation is introduced to a containment building of Shin-Kori NPP which is KSNP (Korean Standard Nuclear Power Plant), to improve its seismic performance. The seismic analysis is performed considering the soil-structure interaction effect by using the linearized model of seismic isolation with SASSI (System for Analysis of Soil-Structure Interaction) program. Finally, the seismic fragility is evaluated based on soil-isolation-structure interaction analysis results.

• 한국가스학회:학술대회논문집
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• 2003.05a
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• pp.165-172
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• 2003

• 한국가스학회:학술대회논문집
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• 2003.05a
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• pp.173-181
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• 2003

• KSCE Journal of Civil and Environmental Engineering Research
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• v.31 no.5A
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• pp.369-378
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• 2011

In this paper, the response analysis of RC(Reinforced Concrete), SC(Steel-Plate Concrete) containment buildings subjected to a large aircraft impact is performed using Autodyn-3D as Hydrocode. Until now, the impact load in the analysis of aircraft impacts has been applied to target structures at the local area by using the impact load-time history function of Riera. However in this paper, the results of aircraft crash are analyzed by using an aircraft model similar to Boeing 767 and verified by comparing the generated history of the aircraft crash against the rigid target with another history by using the Riera's function. To estimate the resistivity of the impact, the response and safety of SC containment buildings, this study is performed by comparing the four cases of plane concrete, reinforced concrete, bonded containment liner plate at reinforced concrete, and SC structure. Thus, the different behaviors between SC and RC structures when they are subjected to the extreme impact load could be anticipated. Consequently, the improved safety is expected by replacing RC structure with SC structure for nuclear power plants.

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

• Journal of the Korean Institute of Gas
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• v.18 no.4
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• pp.8-12
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• 2014

The aim of this study is achieving customer satisfaction by finding cost-effective installation of storage tanks in the gas filling stations. Pros and cons were measured by comparison between above and underground tanks using Pugh methods. In particular, severe problems regarding gas safety are expected to be resolved by reducing the frequency of gas accidents occurred in mass storage facilities. To this end, we have used Pugh's method to compare the advantages and disadvantages of the existing storage tanks-Ground above type and underground type. After the analyses, it is shown that underground containment storage tank can provide the highest degree of safety and Real estate among Ground above type, Underground buried type and underground containment type storage tanks.

• 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.

• International Journal of Naval Architecture and Ocean Engineering
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• v.8 no.6
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• pp.537-553
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• 2016

This paper addresses the safety of two-row tank design by performing the extensive sloshing model tests. Owing to the uncertainties entangled with the scale law transforming the measured impact pressure up to the full scale one, so called comparative approach was taken to derive the design sloshing load. The target design vessel was chosen as 230 K LNG-FPSO with tow-row tank arrangement and the reference vessel as 138 K conventional LNG carrier, which has past track record without any significant failure due to sloshing loads. Starting with the site-specific metocean data, ship motion analysis was carried out with 3D diffraction-radiation program, then the obtained ship motion data was used as 6DOF tank excitation for subsequent sloshing model test and analysis. The statistical analysis was carried out with obtained peak data and the long-term sloshing load was determined out of it. It was concluded that the normalized sloshing impact pressure on 230 K LNG-FPSO with two-row tank arrangement is higher than that of convectional LNG carrier, hence requires the use of reinforced cargo containment system for the sake of failure-free operation without filling limitation.