• Nuclear Engineering and Technology
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• v.15 no.2
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• pp.123-134
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• 1983

Estimates of the rate of heat removal from the containment atmosphere following a loss of colant accident (LOCA) are important to the prediction of containment peak pressure and temperature which are essential parameters in designing the containment building. An overall survey and discussion of mechanisms, experimental results, empirical correlations and analytical models that are relevant to the heat transfer inside the containment have been made. As a result of this review, the current state of the knowledge about tile containment heat transfer can be understood and it is known that more investigations are needed to avoid the misuse of various correlations.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.23 no.2
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• pp.235-243
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• 2010

Three-dimensional structural analysis system for nuclear containment building is presented in this paper. This system includes high-performance plate/shell elements as finite element library. It also adopts numerical modeling technique for unbonded tendon as well as bonded tendon in prestressed concrete structures. This system is constructed by connecting several in-house program to a commercial program DIANA, and then is capable of performing nonlinear analysis for ultimate pressure capacity of nuclear containment building. Finally, three-dimensional structural analysis of CANDU-type containment building is carried out in order to test the reliability of this system. These numerical results are compared with reference values, which obtained from axisymmetric structural analysis.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.16 no.1
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• pp.37-41
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• 2020

The Nuclear containment building is a main safety-related structure that performs shielding and conservation functions to prevent highly radioactive materials from leakage to the outside environment in the case of various environmental conditions and postulated accidents. The containment building contains a reactor, steam generator, pressurizer, tank, reactor coolant system, auxiliary system and engineering safety system, and is designed so that highly radioactive materials above the limits specified in 10 CFR 100 do not escape to the outside environment in the case of LOCA(Loss of Coolant Accident) for instance. The containment metal liner plate(CLP) is a carbon steel plate with a nominal plate thickness of 6 mm, which functions as a mold for the wall and dome of the containment building when concrete is filled, fulfills airtightness to prevent leakage of seriously radioactive materials. In recent years, backside corrosion was found on the liner plate in some domestic nuclear power plants. The main cause of backside corrosion was unfilled concrete. In this paper, an inspection technique of assessing filling suitability for CLP backside concrete is developed. Results show that the validity of inspection technique for CLP backside concrete using vibration sensor is successfully verified.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1231-1242
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• 2020

The Kori Unit 1 will be decommissioned after a permanent shutdown in June 2017. South Korea has a 0.1 mSv/yr exposure limit standard for limited or unlimited site release. This is South Korea's first commercial NPP; therefore, if the containment building is reused as a memorial hall, it will contribute to the improvement of public understanding and enhance the public's acceptance of NPPs. Also, existing Kori Unit 1 nuclear power plant manpower resources can be reused after decommissioning and resident staff and memorial hall visitors can activate nearby commercial areas. Therefore, such a reuse scenario may also prevent an economic recession. The exposure dose was calculated using the following scenarios: worker in the containment building, visitor in the containment building, and worker in buildings other than the containment building. The exposure dose in the buildings was calculated by the RESRAD-BUILD developed by the Argonne National Laboratory (ANL). The preliminary exposure dose and derived concentration guideline level (DCGL) were derived.

• 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.

• Nuclear Engineering and Technology
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• v.55 no.5
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• pp.1742-1756
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• 2023

The hydrogen behavior in a nuclear containment vessel is a significant issue when discussing the potential of hydrogen combustion during a severe accident. After the Fukushima-Daiichi accident in Japan, we have investigated in-depth the hydrogen transport mechanisms by utilizing experimental and numerical approaches. Computational fluid dynamics is a powerful tool for better understanding the transport behavior of gas mixtures, including hydrogen. This paper describes a Large-eddy simulation of gas mixing driven by a high-buoyancy flow. We focused on the interaction behavior of heat and mass transfers driven by the horizontal high-buoyant flow during density stratification. For validation, the experimental data of the Containment InteGral effects Measurement Apparatus (CIGMA) facility were used. With a high-power heater for the gas-injection line in the CIGMA facility, a high-temperature flow of approximately 390 ℃ was injected into the test vessel. By using the CIGMA facility, we can extend the experimental data to the high-temperature region. The phenomenological discussion in this paper helps understand the heat and mass transfer induced by the high-buoyancy flow in the containment vessel during a severe accident.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.975-981
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• 2016

The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2053-2068
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• 2023

Following a loss of coolant accident (LOCA), prestressing concrete containment vessels (PCCVs) may experience high thermal load as well as internal pressure. The high temperature stress would increase the risk of premature damage to the containment, which reduces the safety margin during the increasing internal pressure. However, current investigations cannot clearly address the issues of thermal-pressure coupling effect on damage propagation and thus safety of the containment. Thus, this paper offers three simple and powerful damage parameters to differentiate the severity of damage of the containment. Moreover, despite of the temperature action severely threatening the pressure performance of the containment, the research regarding the improvement of the resistant performance of the containment is quite scarce. Therefore, in this paper, a comprehensive comparison of damage propagation and mechanism between conventional and fiber-reinforced concrete (FRC) containments is performed. The effects of fiber characteristics parameters on damage propagation of structures following the LOCA are also specifically revealed. It is found that the proposed damage indices can properly indicate state of damage in the containment body and the addition of fiber can be used to obviously mitigate the damage propagation in PCCV considering the thermal-pressure coupling.

• Nuclear Engineering and Technology
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• v.11 no.3
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• pp.195-202
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• 1979

In this paper, a cause-consequence chart is applied to evaluate the probability that the containment spray system in a nuclear power plant may not be woring properly, given a demand for spryaing at loss of coolant accident (LOCA). It is shown how the diagram provides a basis for calculating two probability measures for malfunctioning of this system, in case the test policy of the system is taken into account, i.e., average probability that the containment spray cannot be established, and average probability that the containment spray is established : spray stops before the required operating time is over.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.575-602
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• 2009

In the wake of the Three Mile Island accident, vigorous research efforts were initiated to acquire a basic knowledge of the progression and consequences of accidents that involve a substantial degree of core degradation and melting. The primary emphasis of this research was placed on containment integrity, with: i) hydrogen combustion-detonation, ii) steam explosion, iii) direct containment heating (DCH), and iv) melt attack on the BWR Mark-I containment shell identified as energetic processes that could lead to early containment failure (i.e., within the first 24 hours of the accident). Should the core melt fail the reactor vessel, then non-condensable gas production from Molten Core-Concrete Interaction (MCCI) was identified as a mechanism that could fail the containment by pressurization over the long term. One signification question that arose as part of this investigation was the effectiveness of water in terminating an MCCI by flooding the interacting masses from above, thereby quenching the molten core debris and rendering it permanently coolable. Successful quenching of the core melt would prevent basemat melt through, as well as continued containment pressurization by non-condensable gas production, and so the accident progression would be successfully terminated without release of radioactivity to the environment. Based on these potential merits, ex-vessel corium coolability has been the focus of extensive research over the last 20 years as a potential accident management strategy for current plants. In addition, outcomes from this research have impacted the accident management strategies for the Gen III+LWR plant designs that are currently being deployed around the world. This paper provides: i) an historical overview of corium coolability research, ii) summarizes the current status of research in this area, and iii) highlights trends in severe accident management strategies that have evolved based on the findings from this work.