• Nuclear Engineering and Technology
• /
• v.54 no.1
• /
• pp.357-373
• /
• 2022

The high-temperature gas-cooled reactor pebble-bed module project is the first commercial Generation-IV NPP(Nuclear Power Plant) in China. A new joint is used for the vertical support of RPV(Reactor Pressure Vessel). The steel corbel is integrally embedded into the reactor-cabin wall through eight asymmetrically arranged pre-stressed high-strength bolts, achieving the different path transmission of shear force and moment. The vertical monotonic loading test of two specimens is conducted. The results show that the failure mode of the joint is bolt fracture. There is no prominent yield stage in the whole loading process. The stress of bolts is linearly distributed along the height of corbel at initial loading. As the load increases, the height of neutral axis of bolts gradually decreases. The upper and lower edges of the wall opening contact the corbel plate to restrict the rotation of the corbel. During the loading, the pre-stress of some bolts decreases. The increase of the pre-stress strength ratio of bolts has no noticeable effect on the structure stiffness, but it reduces the ultimate bearing capacity of the joint. A simplified calculation model for the elastic stage of the joint is established, and the estimation results are in good agreement with the experimental results.

• Nuclear Engineering and Technology
• /
• v.45 no.4
• /
• pp.553-564
• /
• 2013

This study evaluates the local damage of a turbine in an auxiliary building of a nuclear power plant due to an external impact by using the LS-DYNA finite element program. The wall of the auxiliary building is SC structure and the material of the SC wall plate is high manganese steel, which has superior ductility and energy absorbance compared to the ordinary steel used for other SC wall plates. The effects of the material of the wall, collision speed, and angle on the magnitude of the local damage were evaluated by local collision analysis. The analysis revealed that the SC wall made of manganese steel had significantly less damage than the SC wall made of ordinary steel. In conclusion, an SC wall made of manganese steel can have higher effective resistance than an SC wall made of ordinary steel against the local collision of an airplane engine or against a turbine impact.

• Journal of the Korea Concrete Institute
• /
• v.22 no.6
• /
• pp.817-824
• /
• 2010

The concrete containment structures have been widely used in nuclear power plants, LNG storage tanks, etc., due to their high safety and economic efficiency. The containment structure consists of a bottom slab, wall, ring beam and dome. The shape of the roof dome has a very significant effect on structural safety, the quantity of materials, and constructability; the thickness and curvature of the dome should therefore be determined to give the optimum design. The ring beam plays the role as supports for the dome, resulting in a minimized deformation of the wall. The main issues in designing the ring beam are the correct dimensions of the section and the prestress level. In this study, an efficient design procedure is proposed that can be used to determine an optimal shape and prestress level of the dome and ring beam. In the preliminary design stage of the procedure, the membrane theory of shells of revolution is adopted to determine several plausible alternatives which can be obtained even by hand calculation. Based on the proposed procedures, domes and ring beams of the existing domestic containment structures are analyzed and some improvements are discussed.

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

• Nuclear Engineering and Technology
• /
• v.52 no.12
• /
• pp.2836-2846
• /
• 2020

During severe nuclear power plant (NPP) accidents, a H₂/CO mixture can be generated in the reactor pressure vessel by core degradation and in the containment as well by molten corium-concrete interaction. In spite of its importance, a state-of-the-art methodology predicting H₂/CO combustion risk relies predominantly on empirical correlations. It is therefore necessary to develop a proper methodology for flammability evaluation of H₂/CO mixtures at ex-vessel phases characterized by three factors: CO concentration, high temperature, and diluents. The developed methodology adopted Le Chatelier's law and a calculated non-adiabatic flame temperature model. The methodology allows the consideration of the individual effect of the heat transfer characteristics of hydrogen and carbon monoxide on low flammability limit prediction. The accuracy of the developed model was verified using experimental data relevant to ex-vessel phase conditions. With the developed model, the prediction accuracy was improved substantially such that the maximum relative prediction error was approximately 25% while the existing methodology showed a 76% error. The developed methodology is expected to be applicable for flammability evaluation in chemical as well as NPP industries.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.34 no.2
• /
• pp.119-127
• /
• 2014

An epoxy coating applied to the concrete surface of a containment building deteriorates in hazardous environments such as those containing radiation, heat, and moisture. Unlike metals, the epoxy coating on a concrete liner absorbs and discharges moisture during the degradations process, so it has a different density and volume during service. In this study, acoustic impedance was adopted for characterizing the degradation of a glass-reinforced epoxy coating using the acoustic reflection coefficient (reflectance) on a rough epoxy coating. For estimating the acoustic reflectance on a wavy epoxy coating surface, a probabilistic model was developed to represent the multiple irregular reflections of the acoustic wave from the wavy surface on the basis of the simulated annealing technique. A number of epoxy-coated concrete specimens were prepared and exposed to accelerated aging conditions to induce an artificial aging degradation in them. The acoustic impedance of the degraded epoxy coating was estimated successfully by minimizing the error between a waveform calculated from the mathematical model and a waveform measured from the surface of the rough coating.

• Journal of the Korean Recycled Construction Resources Institute
• /
• v.9 no.4
• /
• pp.451-459
• /
• 2021

The purpose of this study is to evaluate the performance of a solidifying agent for recycling the fine powder separated from the nuclear power plant decommissioned concrete as a solidifying agent(SA) for radioactive waste. In order to evaluate the performance of the solidifying agent, a powder simulating the fine powder of waste concrete separated from the dismantled concrete of a nuclear power plant was produced, and the main variables were the type of binder and the replacement ratio of zeolite. The solidifying agent was evaluated for fluidity performance, compressive strength, and leaching resistance to non-radioactive cesium. The compressive strength of SA increased as the zeolite replacement ratio increased, and the SA containing 5% or more of zeolite showed a compressive strength that was 1.4 to 1.7 times higher than the acceptance criteria. The cesium leaching index of all specimens was 6 or higher, satisfying the acceptance criteria, and the leaching index of SA was 1.47~1.63 times higher than that of OPC. In particular, the average leaching index after 28 days of the 5% zeolite-substituted solidifying agent was 9.15, which was improved by about 6.4% compared to OPC, and it was confirmed that the zeolite was effective in improving the leaching resistance to cesium ions by showing stable performance over the entire period.

• Journal of the Computational Structural Engineering Institute of Korea
• /
• v.27 no.5
• /
• pp.437-450
• /
• 2014

In this study, we developed a methodology to determine the reference parameter for an aircraft impact induced risk assessment of nuclear power plant (NPP) using finite element impact analysis of containment building. The target structure used to develop the method of reference parameter selection is one of the typical Korean PWR type containment buildings. We composed a three-dimensional finite element model of the containment building. The concrete damaged plasticity model was used for the concrete material model. The steels in the tendon, rebar, and liner were modeled using the piecewise-linear stress-strain curves. To evaluate the correlations between structural response and each candidate parameter, we developed Riera's aircraft impact force-time history function with respect to the variation of the loading parameters, i.e., impact velocity and mass of the remaining fuel. For each force-time history, the type of aircraft is assumed to be a Boeing 767 model. The variation ranges of the impact velocity and remaining fuel percentage are 50 to 200m/s, and 30 to 90%, respectively. Four parameters, i.e., kinetic energy, total impulse, maximum impulse, and maximum force are proposed for candidates of the reference parameter. The wellness of the correlation between the reference parameter and structural responses was formulated using the coefficient of determination ($R^2$). From the results, we found that the maximum force showed the highest $R^2$ value in most responses in the materials. The simplicity and intuitiveness of the maximum force parameter are also remarkable compared to the other candidate parameters. Therefore, it can be concluded that the maximum force is the most proper candidate for the reference parameter to assess the aircraft impact induced risk of NPPs.

• Nuclear Engineering and Technology
• /
• v.54 no.8
• /
• pp.2960-2973
• /
• 2022

Since Fukushima nuclear power plant (NPP) accident in 2011, the importance of research on various severe accident phenomena has been emphasized. Particularly, detailed analysis of combustion risk is necessary following the containment damage caused by combustion in the Fukushima accident. Many studies have been conducted to evaluate the risk of local hydrogen concentration increases and flame propagation using computational code. In particular, the potential for combustion by local hydrogen concentration in specific areas within the containment has been emphasized. In this study, the process of flame propagation generated inside a reactor core to containment during a loss of coolant accident (LOCA) was analyzed using MELCOR 2.1 code. Later in the LOCA scenario, it was expected that hydrogen combustion occurred inside the reactor core owing to oxygen inflow through the cold leg break area. The main driving force of the oxygen intrusion is the elevated containment pressure due to the molten corium-concrete interaction. The thermal and mechanical loads caused by the flame threaten the integrity of the containment. Additionally, the containment spray system effectiveness in this situation was evaluated because changes in pressure gradient and concentrations of flammable gases greatly affect the overall behavior of ignition and subsequent containment integrity.

• Journal of Radiation Protection and Research
• /
• v.21 no.4
• /
• pp.313-327
• /
• 1996

As an at-reactor(AR) storage method o( spent fuel, there are horizontal concrete module type, metal storage cask type, concrete storage cask type, dual purpose (transportation and storage) cask type and multi-purpose (transportation, storage and disposal) cask type. All other types except multi-purpose one have been already used for AR dry storage of spent fuels after obtaining operation license in various foreign countries. Also the development of multi-purpose type has been continued for operation license. In America, Japan, Germany, Canada, Spain, Switzerland, and Czech Republic, etc., AR dry storage facilities are under operation or on propulsion, and spent fuels are transported to interim storage facility or reprocessing plant after dry storage at reactor temporarily. At Wolsung site, in case of Korea, concrete silo type has already been introduced, and it is believed to be inevitable to store spent fuels at reactor temporarily, considering the reality that storage capacity of spent fuel is approaching to the limit in some nuclear power plants. In this report, the system characteristics, design requirements, technical standards and status of AR storage system, which is suitable for domestic site such as Kori, have been studied. In most cases, the licensed period of storage cask is limited up to 20 years and the integrity of material and maintenance of leaktightness are required during the whole service life.