• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2069-2087
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• 2023

In nuclear power plant (NPP) accidents, the containment is subject to high temperatures and high internal pressures, which may further trigger serious chain accidents such as core meltdown and hydrogen explosion, resulting in a significantly higher accident level. Therefore, studying the mechanical performance of a containment under high temperature and high internal pressure is relevant to the safety of NPPs. Based on similarity principles, the 1:3.2 scale model of a prestressed concrete containment vessel (PCCV) of a NPP was designed. The loading method, which considers the thermal-pressure coupling conditions, was used. The mechanical response of the PCCV was investigated with a simultaneous increase in internal pressure and temperature, and the failure mechanism of the PCCV under thermal-pressure coupling conditions was revealed.

• Journal of the Korea Institute of Information and Communication Engineering
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• v.25 no.1
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• pp.102-107
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• 2021

This paper discusses the containment control problem for multi-agent systems with input saturations. The goal of the containment control is to obtain swarming behavior by driving follower agents into the convex hull which is spanned by multiple leader agents. This paper considers multiple leader agents moving at the same constant speed. Then, to solve the containment problem for moving leaders, we propose a PI-based distributed control algorithm. We next analyze the convergence of follower agents to the desired positions. Specifically, we apply the integral-type Lyapunov function to take into account the saturation nonlinearity. Then, based on Lasalle's Invariance Principle, we show that the asymptotic convergence of error states to zero for any positive constant gains. Finally, numerical examples with the static and moving leaders are provided to validate the theoretical results.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1140-1153
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• 2016

Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA) by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model). In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP) containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code's results.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2002.04a
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• pp.377-384
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• 2002

In this palter, a finite element technique is applied to both reinforced concrete and prestressed concrete containment vessels to predict the ultimate pressure capacity of the vessels subjected to internal pressure due to accident. The so-called volume-control technique is utilized to control the change in volume enclosed by the cylindrical containment vessels and layered shell elements equipped with a pressure node is utilizing to model the PSC vessels. The finite element analysis is carried out to obtain both global and local failure behavior of prestressed concrete nuclear containment vessels. nalytical results are verified by comparison with experimental data.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.931-942
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• 2019

Radioactive materials are released from the molten core into the containment at the time of a severe accident in a nuclear power plant (NPP). Filtered containment venting system is a popular and effective safety measure installed to obstruct the uncontrolled escape of radioactive materials due to the over pressurization of the containment. Different designs of filtered containment venting system (FCVS) are available today, each being the result of extensive research and development varying in one way or the other. This paper gives an elaborate description of the different types of FCVS currently being used, the current usage status in over 17 countries and the legislations regarding it. The recent researches being carried out in this field has also been discussed in detail. This present paper focuses on the critical review of existing FCVS, reports the challenges faced by it and highlights the potential developments to overcome the difficulties.

• Journal of the Korea Institute of Information and Communication Engineering
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• v.27 no.1
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• pp.109-116
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• 2023

In this paper, we study the containment control problem for second-order multi-agent systems, which consists of multiple leaders and followers. The goal is to drive the followers toward the convex hull spanned by the leaders. Thus, the swarm behavior can be obtained by controlling the entire group by the leaders. This paper considers the leaders move at a constant speed and the followers have input saturations. Moreover, we assume that the followers can exchange information with neighbors, and only relative state information is available. Under these assumptions, we propose the Proportional-Integral based distributed control algorithm to solve the containment control problem with moving leaders. Moreover, based on Lasalle's invariance principle, the conditions for the control gains that guarantee the convergence of the followers to the convex hull spanned by the leaders are investigated, and it was shown that it can be designed only using the system parameter. Finally, the simulations are conducted to validate the theoretical result.

• Journal of the Earthquake Engineering Society of Korea
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• v.25 no.4
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• pp.179-189
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• 2021

Conditional spectra (CS) are applied to the seismic fragility assessment of a nuclear power plant (NPP) containment building for comparison with a relevant conventional uniform hazard response spectrum (UHRS). Three different control frequencies are considered in developing conditional spectra. The contribution of diverse magnitudes and epicentral distances is identified from deaggregation for the UHRS at a control frequency and incorporated into the conditional spectra. A total of 30 ground motion records are selected and scaled to simulate the probability distribution of each conditional spectra, respectively. A set of lumped mass stick models for the containment building are built considering nonlinear bending and shear deformation and uncertainty in modeling parameters using the Latin hypercube sampling technique. Incremental dynamic analysis is conducted for different seismic input models in order to estimate seismic fragility functions. The seismic fragility functions and high confidence of low probability of failure (HCLPF) are calculated for different seismic input models and analyzed comparatively.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.793-803
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• 2021

The small PWR has been paid more and more attention due to its diversity of application and flexibility in the site selection. However, the large core power density, the small containment space and the rapid accident progress characteristics make it difficult to control the containment pressure like the traditional PWR during the LOCA. The pressure suppression system has been used by the BWR since the early design, which is a suitable technique that can be applied to the small PWR. Since the configuration and operating conditions are different from the BWR, the pressure suppression system should be redesigned for the small PWR. Conducting the experiments on the scale down test facility is a good choice to reproduce the prototypical phenomena in the test facility, which is both economical and reasonable. A systematic scaling method referring to the H2TS method was proposed to determine the geometrical and thermohydraulic parameters of the pressure suppression containment response test facility for the small PWR conceptual design. The containment and the pressure suppression system related thermohydraulic phenomena were analyzed with top-down and bottom-up scaling methods. A set of the scaling criteria were obtained, through which the main parameters of the test facility can be determined.

• Computers and Concrete
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• v.26 no.5
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• pp.429-438
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• 2020

To shorten the construction times of nuclear facility structures, three high-strength concrete mixtures were developed with specific consideration given to their curing temperatures, their economic efficiency, and the practicality of their quality control. This study was conducted to examine the temperature rise profiles of these three concrete mixtures and the potential for early-age thermal cracking in the primary containment vessel of a nuclear reactor with a wall thickness of 1200 mm. The one-layer placement height of the concrete for the primary containment vessel was increased from the conventional 3 m to 3.5 m. A nonlinear finite element analysis (FEA) was conducted using the thermal properties of concrete determined from the isothermal hydration and adiabatic hydration tests, and tuned through comparisons made with temperature rise profiles obtained for 1200-mm-thick mock-up wall specimens cured at temperatures of 5, 20, and 35℃. The hydration heat performance of the three concrete mixtures and their potential to produce thermal cracking in nuclear facilities indicate that the mixtures have considerable potential for practical application to the primary containment vessel of a nuclear reactor at various curing temperatures, fulfilling the minimum requirements of the ACI 301 and minimizing the likelihood of the occurrence of thermal cracks.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2012.11a
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• pp.165-166
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• 2012

A thermal stresses by heat of hydration was analyzed according to a change of a pour height in reactor containment building. In case of more than 3.6m pouring height a crack index by heat of hydration analysis resulted in less than 1 because there is not a construction joint of vertical direction and for a self-restraint effect of circumferential section shape. Therefore detailed consideration on a mixture proportion of binder type, quantity in concrete and selection of a form in seasonal air temperature is needed for a control of tensile stress by heat of hydration.