• Nuclear Engineering and Technology
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• v.35 no.2
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• pp.154-164
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• 2003

The explicit consideration of different pressurization rates in estimating the probabilities of containment failure modes has a profound effect on the confidence of containment performance evaluation that is so critical for risk assessment of nuclear power plants. Except for the sophisticated NUREG-1150 study, many of the recent containment performance analyses (through Level 2 PSAs or IPE back-end analyses) did not take into account an explicit distinction between slow and fast pressurization in their analyses. A careful investigation of both approaches shows that many of the approaches adopted in the recent containment performance analyses exactly correspond to the NUREG-1150 approach for the prediction of containment failure mode probabilities in the presence of fast pressurization. As a result, it was expected that the existing containment performance analysis results would be subjected to greater or less conservatism in light of the ultimate failure mode of the containment. The main purpose of this paper is to assess potential conservatism of a plant-specific containment performance analysis result in light of containment failure mode probabilities.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2499-2508
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• 2021

Passive containment cooling systems (PCCSs) have been actively studied to improve the inherent safety of nuclear power plants. Hered, we present two concepts, open-loop PCCS (OL-PCCS) and closed-loop PCCS (CL-PCCS), applicable to the PWR with a concrete-type containment. We analyzed the heat-removal performance and flow instability of these PCCS concepts using the GOTHIC code. In both cases, PCCS performance improved when a passive containment cooling heat exchanger (PCCX) was installed in the lower part of the containment building. The OL-PCCS was found to be superior in terms of heat-removal performance. However, in terms of flow instability, the OL-PCCS was more vulnerable than the CL-PCCS. In particular, the possibility of flow instability was higher when the PCCX was installed in the upper part of the containment. Therefore, the installation location of the OL-PCCS should be restricted to minimize flow instability. Conversely, a CL-PCCS can be installed without any positional restriction by adjusting the initial system pressure within the loop, which eliminates flow instability. These results could be used as base data for the thermo-hydraulic evaluation of PCCS in PWR with a large dry concrete-type containment.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.21 no.1
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• pp.409-416
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• 2020

The seismic performance of a containment structure should be secured to maintain the structural soundness of a containment structure under various earthquakes that occur globally. Therefore, an analysis of the seismic performance of a modular containment structure for a small modular reactor is also required. To analyze the seismic performance of modular containment, FEM models with contact surfaces between the modules and tendon were prepared and the modal and seismic analyses were performed. The displacement, stress, and gap size of modular containment under earthquake wave were analyzed. The effects of the tendon force, friction coefficient, and earthquake wave on the seismic performance were analyzed. The seismic performance of monolithic containment was also analyzed for comparison. In the 1st and 2nd natural modes, which most likely affect, the modular containment showed horizontal dynamic behavior, which is similar to monolithic containment, because of the combined effects of the tendon force and friction force between modules. When the combined effect is sufficient, the seismic performance of the modular containment is secured over a certain level. An additional increase in seismic performance is expected when some material with a larger friction coefficient is adopted on the contact surface.

• Nuclear Engineering and Technology
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• v.53 no.4
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• pp.1134-1145
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• 2021

A design parameter study is presented for the closed-loop type passive containment cooling system (PCCS) which is equipped with two heat exchangers: one installed at the inside of the containment and the other submerged in the water pool at the outside of the containment. A GOTHIC code model for PCCS performance analyses was set up and the design parameters such as the heat exchanger sizes, locations, and water pool tank volumes were analyzed to investigate the feasibility of installing this type of PCCS in PWRs like OPR-1000 being operated in Korea. We identified the size of the circulation loop and heat exchangers as major design parameters affecting the performance of PCCS. The analyses showed that the heat exchangers in the inside of the containment would be more influential on the heat removal capability of PCCS than that installed in the water pool at the outside of the containment. Hence, it was recommended to down-size the heat exchangers in the water pool to optimize PCCS without compromising its performance. Based on the parametric study, it was demonstrated that a closed-loop type PCCS could be designed sufficiently compact for installation in the available space within the containment of PWRs like OPR-1000.

• Nuclear Engineering and Technology
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• v.49 no.5
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• pp.1079-1089
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• 2017

Nuclear power plants under expansion and under construction in China are mostly located in coastal areas, which means they are at risk of suffering strong earthquakes and subsequent tsunamis. This paper presents a safety analysis for a new reinforced concrete containment vessel in such events. A finite element method-based model was built, verified, and first used to understand the seismic performance of the containment vessel under earthquakes with increased intensities. Then, the model was used to assess the safety performance of the containment vessel subject to an earthquake with peak ground acceleration (PGA) of 0.56g and subsequent tsunamis with increased inundation depths, similar to the 2011 Great East earthquake and tsunami in Japan. Results indicated that the containment vessel reached Limit State I (concrete cracking) and Limit State II (concrete crushing) when the PGAs were in a range of 0.8-1.1g and 1.2-1.7g, respectively. The containment vessel reached Limit State I with a tsunami inundation depth of 10 m after suffering an earthquake with a PGA of 0.56g. A site-specific hazard assessment was conducted to consider the likelihood of tsunami sources.

• Journal of the Earthquake Engineering Society of Korea
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• v.28 no.4
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• pp.223-231
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• 2024

For the OPR1000, a standard power plant in Korea, an analytical model of the containment building considering voids and deterioration was built with multilayer shell elements. Voids were placed in the vulnerable parts of the analysis model, and the deterioration effects of concrete and rebar were reflected in the material model. To check the impact of voids and deterioration on the seismic performance of the containment building, iterative push-over analysis was performed on four cases of the analytical model with and without voids and deterioration. It was found that the effect of voids with a volume ratio of 0.6% on the seismic performance of the containment building was insignificant. The effect of strength reduction and cross-sectional area loss of reinforcement due to deterioration and the impact of strength increase of concrete due to long-term hardening offset each other, resulting in a slight increase in the lateral resistance of the containment building. To determine the limit state that adequately represents the seismic performance of the containment building considering voids and deterioration, the Ogaki shear strength equation, ASCE 43-05 low shear wall allowable lateral displacement ratio, and JEAC 4601 shear strain limit were compared and examined with the analytically derived failure point (ultimate point) in this study.

• Nuclear Engineering and Technology
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• v.28 no.2
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• pp.129-136
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• 1996

Influence diagram method is applied to containment performance analysis of Young-Gwang 3&4 in an effort to overcome some drawbacks of current containment performance analysis method. Event tee methodology has been adopted as a containment performance analysis method. There are, however, some drawbacks on event tree methodology. This study is to overcome three major drawbacks of the current containment performance analysis method : 1) Event tree cannot express dependency between events explicitly. 2) Accident Progression Event Tree (APET) cannot represent entire containment system. 3) It is difficult to consider decision making problem. To resolve these problems, influence diagrams, is proposed. In the present ok, the applicability of the influence diagrams has been demonstrated for YGN 3&4 containment performance analysis and accident management strategy assessments of this study are in good agreement with those of YGN 3&4 IPE. Sensitivity analysis has been peformed to identify relative important variables for each early containment failure, late containment and basemat melt-though. In addition, influence diagrams are used to assess two accident management strategies : 1) RCS depressurization, 2) cavity flooding. It is shown that influence diagrams can be applied to the containment performance analysis.

• 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.49 no.5
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• pp.941-952
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• 2017

Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS) of an innovative pressurized water reactor (PWR). A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT) geometry, PCCS heat exchanger (PCCX) location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1513-1525
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• 2024

An improved mitigation system for thermally-induced steam generator tube rupture accidents was introduced to prevent direct environmental release of fission products bypassing the containment in the OPR1000. This involves injecting bypassed steam into the containment, cooling, and decontaminating it using a water coolant tank. To evaluate its performance, a severe accident analysis was performed using the MELCOR 2.2 code for OPR1000. Simulation results show that the proposed system sufficiently prevented the release of radioactive nuclides (RNs) into the environment via containment injection. The pool scrubbing system effectively decontaminated the injected RN and consequently reduced the aerosol mass in the containment atmosphere. However, the decay heat of the collected RNs causes re-vaporization. To restrict the re-vaporization, an external water source was considered, where the decontamination performance was significantly improved, and the RNs were effectively isolated. However, due to the continuous evaporation of the feed water caused by decay heat, a substantial amount of steam is released into the containment. Despite the slight pressurization inside the containment by the injected and evaporated steam, the steam decreased the hydrogen mole fraction, thereby reducing the possibility of ignition.