• Nuclear Engineering and Technology
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• 제47권7호
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• pp.884-894
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• 2015

Background: Fibers in concrete resist the growth of cracks and enhance the postcracking behavior of structures. The addition of fibers into a conventional reinforced concrete can improve the structural and functional performance of safety-related concrete structures in nuclear power plants. Methods: The influence of fibers on the ultimate internal pressure capacity of a prestressed concrete containment vessel (PCCV) was investigated through a comparison of the ultimate pressure capacities between conventional and fiber-reinforced PCCVs. Steel and polyamide fibers were used. The tension behaviors of conventional concrete and fiber-reinforced concrete specimens were investigated through uniaxial tension tests and their tension-stiffening models were obtained. Results: For a PCCV reinforced with 1% volume hooked-end steel fiber, the ultimate pressure capacity increased by approximately 12% in comparison with that for a conventional PCCV. For a PCCV reinforced with 1.5% volume polyamide fiber, an increase of approximately 3% was estimated for the ultimate pressure capacity. Conclusion: The ultimate pressure capacity can be greatly improved by introducing steel and polyamide fibers in a conventional reinforced concrete. Steel fibers are more effective at enhancing the containment performance of a PCCV than polyamide fibers. The fiber reinforcementwas shown to bemore effective at a high pressure loading and a lowprestress level.

• Nuclear Engineering and Technology
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• 제53권6호
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• pp.1756-1768
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• 2021

The containment filtered venting system (CFVS) filters the atmosphere of the containment building and discharges a part of it to the outside environment to prevent containment overpressure during severe accidents. The Korean CFVS has a tank that filters fission products from the containment atmosphere by pool scrubbing, which is the primary decontamination process; however, prediction of its performance has been done based on researches conducted under mild conditions than those of severe accidents. Bubble behavior in a pool is a key parameter of pool scrubbing. Therefore, the bubble behavior in the pool was analyzed under various injection flow rates observed at the venturi nozzles used in the Korean CFVS using a wire-mesh sensor. Based on the experimental results, void fraction model was modified using the existing correlation, and a new bubble size prediction model was developed. The modified void fraction model agreed well with the obtained experimental data. However, the newly developed bubble size prediction model showed different results to those established in previous studies because the venturi nozzle diameter considered in this study was larger than those in previous studies. Therefore, this is the first model that reflects actual design of a venturi scrubbing nozzle.

• 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.

• Nuclear Engineering and Technology
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• 제55권1호
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• pp.144-155
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• 2023

The core catcher is used as a passive safety system in new generation nuclear power plants to create a space in the containment for the placing and cooling of the molten corium under various severe accidents. This research investigates the role of the core catcher in the VVER-1000 reactor containment system in mitigating the effects of core meltdown under various severe accidents within the context of the Ex-vessel Melt Retention (EVMR) strategy. Hence, a comparison study of three severe accidents is conducted, including Station Black-Out (SBO), SBO combined with the Large Break Loss of Coolant Accident (LB-LOCA), and SBO combined with the Small Break Loss of Coolant Accident (SB-LOCA). Numerical comparative simulations are performed for the aforementioned scenario with and without the EX-vessel core-catcher. The results showed that considering the EX-Vessel core catcher reduces the amount of hydrogen by about 18.2 percent in the case of SBO + LB-LOCA, and hydrogen production decreases by 12.4 percent in the case of SBO + SB-LOCA. Furthermore, in the presence of an EX-Vessel core-catcher, the production of gases such as CO and CO2 for the SBO accident is negligible. It was revealed that the greatest decrease in pressure and temperature of the containment is related to the SBO accident.

• Nuclear Engineering and Technology
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• 제56권2호
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• pp.611-623
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• 2024

The ISP-47 TOSQAN experiment was analyzed with containmentFOAM which is an open-source CFD code based on OpenFOAM. The containment phenomena taking place during the experiment are gas mixing, stratification and wall condensation in a mixture composed of steam and non-condensable gas. The k-ω SST turbulence model was adopted with buoyancy turbulence models. The wall condensation model used is based on the diffusion layer approach. We have simulated the full TOSQAN experiment which had a duration 20000 s. Sensitivity studies were conducted for the buoyancy turbulence models with SGDH and GGDH and there were not significant differences. All the main features of the experiments namely pressure history, temperature, velocity and gas species evolution were well predicted by containemntFOAM. The simulation results confirmed the formation of two large flow stream circulations and a mixing zone resulting by the combined effects of the condensation flow and natural convection flow. It was found that the natural convection in lower region of the vessel devotes to maintain two large circulations and to be varied the height of the mixing zone as result of sensitivity analysis of non-condensing wall temperature. The computational results obtained with the 2D mesh grid approach were comparable to the experimental results.

• Nuclear Engineering and Technology
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• 제56권5호
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• pp.1638-1653
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• 2024

Given the unpredictability of the occurrence of the earthquake and other potential disasters into consideration, the nuclear power plant may be confronted with beyond design-basis earthquake load in the future. The containment structure may be severely damaged under such severe earthquake loading, increasing the risk of containment concrete cracking and potential radioactive materials leaking. Moreover, initial damage caused by the earthquake may significantly alter the pressure performance of the containment under follow-up internal pressure. To compromise the dangers of beyond design-basis earthquake to the containment, an alternative of replacing the conventional concrete with fiber-reinforced concrete (FRC) to upgrade the seismic resistance capacity of the containment is attempted and thoroughly researched. In this study, the influence of various fiber types such as rigid fiber and mixed fiber is regarded to constitute fiber-reinforced PCCVs. The physical properties of traditional and fiber-reinforced PCCVs under earthquake ground motions are scientifically compared and identified by using traditional and proposed evaluation indices. The results indicate that both the traditional evaluation index (i.e. top displacement, stress, strain) and the proposed damage index are greatly reduced by the practice of fiber strengthening under earthquake ground motions.

• 한국콘크리트학회:학술대회논문집
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• 한국콘크리트학회 2003년도 가을 학술발표회 논문집
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• pp.608-612
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• 2003

In the seismic response analysis, the artificial earthquake time history is generated to do the exact seismic analysis for the complex structural system like as containment building. In the present study the several simulated earthquakes are generated by use of SIMQKE program and the time history dynamic analysis of containment building is performed. Also, the seismic responses are statistically analyzed. The seismic response uncertainty arisen from the simulation of earthquakes is one of major uncertainties and the statistical description is needed to account for the random nature of earthquake.

• 한국지하수토양환경학회:학술대회논문집
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• 한국지하수토양환경학회 2002년도 추계학술발표회
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• pp.24-27
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• 2002

The optimal designs of groundwater problems of contaminant containment and cleanup using linear programming and genetic algorithm are provided. In the containment problem, genetic algorithm shows the superior feature to linear programming. In cleanup problem, genetic algorithm makes reasonable optimal design. Un this study, it is demonstrated through numerical experiments that genetic algorithm can be applied to remedial designs of groundwater problems.

• Nuclear Engineering and Technology
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• 제29권6호
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• pp.444-450
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• 1997

The current ASME code for reinforced concrete containment structures are not based on probability concepts. The stochastic nature of natural hazard or accidental loads and the variations of material properties require a probabilistic approach for a rational assessment of structural safety and performance. The paper develops design load factors for the serviceability limit state of reinforced concrete containment structures. The target limit state probability is determined and the load factors are calculated by the numerical analysis. Design load factors are proposed and carried out the reliability assessments.

• 한국지진공학회:학술대회논문집
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• 한국지진공학회 2003년도 춘계 학술발표회논문집
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• pp.118-125
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• 2003

Seismic damage assessment program for containment structure is developed. The program has been established through the combination of inelastic seismic analysis program and 3-D animation program. Damage indices at finite element level and structural level have been introduced for the seismic damage assessment. The seismic damage assessment program makes it possible to analyze in real-time the actual resistance capacity and damage level of containment structure. It will be expected that the program enables to establish the measures more quickly under the earthquake event.