• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.386-400
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• 2022

Seismic fragility analysis, a part of seismic probabilistic risk assessment (SPRA), is commonly used to establish the relationship between a representative property of earthquakes and the failure probability of a structure, component, or system. Current guidelines on the SPRA of nuclear power plants (NPPs) used worldwide mainly reflect the earthquake characteristics of the western United States. However, different earthquake characteristics may have a significant impact on the seismic fragility of a structure. Given the concern, this study aimed to investigate the effects of earthquake characteristics on the seismic fragility of concrete containments housing the OPR-1000 reactor. Earthquake time histories were created from 30 ground motions (including those of the 2016 Gyeongju earthquake) by spectral matching to the site-specific response spectrum of Hanbit nuclear power plants in South Korea. Fragility curves of the containment structure were determined under the linear response history analysis using a lumped-mass stick model and 30 ground motions, and were compared in terms of earthquake characteristics. The results showed that the median capacity and high confidence of low probability of failure (HCLPF) tended to highly depend on the sustained maximum acceleration (SMA), and increase when using the time histories which have lower SMA compared with the others.

• Nuclear Engineering and Technology
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• v.55 no.12
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• pp.4504-4517
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• 2023

After a loss of coolant accident (LOCA) in the prestressed concrete containment vessels (PCCVs) of nuclear power plants, the coupling of temperature and pressure can significantly affect the mechanical properties of the PCCVs. However, there is no consensus on how this coupling affects the failure mechanism of PCCVs. In this paper, a simplified finite element modeling method is proposed to study the effect of temperature and pressure coupling on PCCVs. The experiment results of a 1:4 scale PCCV model tested at Sandia National Laboratory (SNL) are compared with the results obtained from the proposed modeling approach. Seven working conditions are set up by varying the internal and external temperatures to investigate the failure mechanism of the PCCV model under the coupling effect of temperature and pressure. The results of this paper demonstrate that the finite element model established by the simplified finite element method proposed in this paper is highly consistent with the experimental results. Furthermore, the stress-displacement curve of the PCCV during loading can be divided into four stages, each of which corresponds to the damage to the concrete, steel liner, steel rebar, and prestressing tendon. Finally, the failure mechanism of the PCCV is significantly affected by temperature.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1396-1408
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• 2020

Filtered containment venting system (FCVS) is one of the severe accident mitigation systems designed to release containment pressurization to maintain its integrity. The thermal-hydraulic behaviors in FCVSs are important since they affect the operation characteristics of the FCVS. In this study, a representative FCVS was modeled by RELAP5/Mod3.3 code, and the Station BlackOut (SBO) was chosen as an accident scenario. The thermal-hydraulic behaviors of an FCVS during long-term operation with two venting strategies (open-and-close strategy, open-and-non-close strategy) and the sensitivity analysis of important parameters were investigated. The results show that the FCVS can operate up to 250 h with a periodic open-and-close strategy during an SBO. Under the combined effects of steam condensation and water evaporation, the solution inventory in the FCVS increases during the venting phase and decreases during the intermission phase, showing a periodic pattern. Under this condition, the appropriate initial water level is 3-4 m; however, it should be adjusted according to the environment temperature. The FCVS can accommodate a decay heat power of 150-260 kW and may need to feed water for a higher decay heat power or drain water for a lower decay heat power during the late phase. The FCVS can function within an opening pressure range from 450 kPa to 500 kPa and a closing pressure range between 250 kPa and 350 kPa. When the open-and-non-close strategy is adopted, the solution inventory increases quickly in the early venting phase due to steam condensation and then decreases gradually due to the evaporation of water; drying-up may occur in the late venting phase. Decreasing the venting pipe diameter and increasing the initial water level can mitigate the evaporation of the scrubbing solution. These results are expected to provide useful references for the design and engineering application of FCVSs.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.803-811
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• 2024

Nuclear accidents such as Fukushima Daiichi have highlighted the potential of passive safety systems to replace or complement active safety systems as part of the overall prevention and/or mitigation strategies. In addition, passive systems are key features of Small Modular Reactors (SMRs), for which they are becoming almost unavoidable and are part of the basic design of many reactors available in today's nuclear market. Nevertheless, their potential to significantly increase the safety of nuclear power plants still needs to be strengthened, in particular the ability of computer codes to determine their performance and reliability in industrial applications and support the safety demonstration. The PASTELS project (September 2020-February 2024), funded by the European Commission "Euratom H2020" programme, is devoted to the study of passive systems relying on natural circulation. The project focuses on two types, namely the SAfety COndenser (SACO) for the evacuation of the core residual power and the Containment Wall Condenser (CWC) for the reduction of heat and pressure in the containment vessel in case of accident. A specific design for each of these systems is being investigated in the project. Firstly, a straight vertical pool type of SACO has been implemented on the Framatome's PKL loop at Erlangen. It represents a tube bundle type heat exchanger that transfers heat from the secondary circuit to the water pool in which it is immersed by condensing the vapour generated in the steam generator. Secondly, the project relies on the CWC installed on the PASI test loop at LUT University in Finland. This facility reproduces the thermal-hydraulic behaviour of a Passive Containment Cooling System (PCCS) mainly composed of a CWC, a heat exchanger in the containment vessel connected to a water tank at atmospheric pressure outside the vessel which represents the ultimate heat sink. Several activities are carried out within the framework of the project. Different tests are conducted on these integral test facilities to produce new and relevant experimental data allowing to better characterize the physical behaviours and the performances of these systems for various thermo-hydraulic conditions. These test programmes are simulated by different codes acting at different scales, mainly system and CFD codes. New "system/CFD" coupling approaches are also considered to evaluate their potential to benefit both from the accuracy of CFD in regions where local 3D effects are dominant and system codes whose computational speed, robustness and general level of physical validation are particularly appreciated in industrial studies. In parallel, the project includes the study of single and two-phase natural circulation loops through a bibliographical study and the simulations of the PERSEO and HERO-2 experimental facilities. After a synthetic presentation of the project and its objectives, this article provides the reader with findings related to the physical analysis of the test results obtained on the PKL and PASI installations as well an overall evaluation of the capability of the different numerical tools to simulate passive systems.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2014.11a
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• pp.212-213
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• 2014

The purpose of this study is to suggest analysis model of RCB construction in nuclear power plant. For the objective, This study drew the risk factors of RCB construction from existing literature. The results of the study proposed analysis model made hierarchy in rebar, form, and concrete work. These will be baseline data for risk management in construction project of nuclear power plant.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2004.10a
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• pp.312-319
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• 2004

This paper is mai y focused to develop new concrete material model such as ultimate failure surface in compression-compression region, hardening rule and cracking criteria which are basically used in the nonlinear finite element analysis of nuclear prestressed concrete containment building. From the Kepri's experimental results, failure surface of the concrete based on the elasto-plastic material model is modified and new cracking criteria is proposed. Nonlinear FE analysis program using a new material model is implemented to analysis plane concrete. Finally, numerical simulation to compare the performance of the new material model with experimental results is employed. The numerical results by the proposed model in this study agree very well with the experimental data.

• 유체기계공업학회:학술대회논문집
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• 2001.11a
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• pp.404-410
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• 2001

A new thermal type fluid level transmitter was designed and tested at the HITROL R&D institute. The relation of heat transfer and electric resistance was adopted as an operation principle. The length of a fabricated level transmitter was two meters and a water under normal temperature was used as a working fluid for the experiment Finally, the new product could have a high precision, acceptable accuracy and reasonable response time. Foreign-made level transmitter of this type is under in use for measuring water level of containment building in nuclear power plants so far. It is expected that new product will substitute it.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.663-668
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• 1996

A two-dimensional continuum model for the prediction of the hydrogen mixing phenomena in the containment compartment under the severe accident conditions is developed. The model could predict well the distribution of time-dependent hydrogen concentration for selected HEDL Experiment. For a simulation of these experiments, the hydrogen is mixed uniform over the test compartment. To predict the extent of non-uniform distribution, the dominant factors such as the geometrical shape of obstacle and velocity of source injection in mixing phenomena are investigated. If the obstacle disturbing the flow of gas mixture exists in the compartment, the uniform distribution of hydrogen may be not guaranteed. The convective circulation of gas flow is separately formed up and down of the obstacle position, which makes a difference of hydrogen concentration between the upper and lower region of the compartment. The recirculation flow must have a considerable mass flow rate relative to velocity of the source injection to sustain the well-mixed conditions of hydrogen.

• Journal of Advanced Marine Engineering and Technology
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• v.40 no.2
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• pp.112-119
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• 2016

Many studies have been conducted on the performance of a passive autocatalytic recombiner (PAR), but not many have focused on the locations where the PAR is installed. During a severe accident in a nuclear reactor containment, a large amount of hydrogen gas can be produced and released into the containment, leading to hydrogen deflagration or a detonation. A PAR is a hydrogen mitigation method that is widely implemented in current and advanced light water reactors. Therefore, for this study, a PAR was installed at different locations in order to investigate the difference in hydrogen reduction rate. The results indicate that the hydrogen reduction rate of a PAR is proportional to the distance between the hydrogen induction location and the bottom wall.

• Proceedings of the Computational Structural Engineering Institute Conference
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• 2007.04a
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• pp.694-697
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• 2007

In this paper, a nonlinear finite element analysis program NUCAS, which has been developed for assessment of pressure capacity and failure mode for nuclear containment building is described. Degenerated shell element with assumed strain method and low-order solid element with enhanced assumed strain method is adapted to microscopic material and elasto-plastic material model, respectively. Finally, the performance of the developed program is tested and demonstrated with several examples. From the numerical tests, the present results show a good agreement with experimental data or other numerical results.