• Nuclear Engineering and Technology
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• 제55권6호
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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.

• 한국정보통신학회논문지
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• 제25권1호
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• pp.102-107
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• 2021

본 논문에서는 입력 포화가 존재하는 다중 에이전트 시스템의 봉쇄제어 문제를 다룬다. 봉쇄제어의 목표는 추종 에이전트들을 리더 에이전트들에 의해 형성된 convex hull 안으로 몰아넣음으로써 군집 행동을 얻는 것이다. 본 논문에서는 일정한 속도로 움직이는 리더 에이전트들을 고려한다. 움직이는 리더들을 고려한 봉쇄 문제를 해결하기 위하여 PI기반의 분산제어 알고리즘을 제안한다. 다음으로 추종 에이전트들의 목표 위치로의 수렴성을 해석한다. 구체적으로 포화 비선형성을 고려하기 위하여 적분 형태의 리아프노프 함수를 적용한다. 그리고 Lasalle's Invariance Principle을 기반으로 임의의 상수 이득들에 대하여 오차 상태들의 점근적 수렴성을 보인다. 마지막으로 고정된 리더들과 일정한 속도로 움직이는 리더들을 고려한 시뮬레이션을 진행하여 이론적 결과를 검증하였다.

• Nuclear Engineering and Technology
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• 제48권5호
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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.

• 한국전산구조공학회:학술대회논문집
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• 한국전산구조공학회 2002년도 봄 학술발표회 논문집
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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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• 제51권4호
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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.

• 한국정보통신학회논문지
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• 제27권1호
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• pp.109-116
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• 2023

본 논문에서는 다중 리더 에이전트와 추종 에이전트들로 구성된 2차 다중 에이전트 시스템의 봉쇄제어 문제를 연구하였다. 봉쇄제어의 목표는 추종 에이전트들을 다중 리더 에이전트들에 의해 생성되는 convex hulll을 추종하도록 하는 데에 있다. 따라서 리더 에이전트들에 의해 전체 그룹을 제어함으로써 다중 에이전트 시스템의 군집 행동을 얻을 수 있다. 본 논문에서 리더 에이전트들은 일정한 속도로 움직이고 추종 개체들은 입력 포화가 존재하는 경우를 고려하였다. 또한 추종 에이전트들은 이웃한 에이전트들과 상태 정보를 교환할 수 있고, 이웃과의 상대 상태 정보만 이용 가능하다 가정하였다. 이러한 가정하에 움직이는 리더 에이전트들을 고려한 봉쇄제어 문제를 해결하기 위해 비례-적분 기반의 분산제어 알고리즘을 제안하였다. 또한, 라살레 불변의 법칙을 기반으로 추종 에이전트들의 리더 에이전트들에 의해 생성되는 convex hull로 수렴을 보장하는 제어 이득들에 대한 조건들을 조사하였고 시스템 파라미터의 정보만으로 설계할 수 있음을 보였다. 마지막으로 모의실험을 통한 이론적 결과를 검증하였다.

• 한국지진공학회논문집
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• 제25권4호
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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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• 제53권3호
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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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• 제26권5호
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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.

• 한국건축시공학회:학술대회논문집
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• 한국건축시공학회 2012년도 추계 학술논문 발표대회
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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.