• 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.

• Nuclear Engineering and Technology
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• 제53권5호
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• pp.1540-1555
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• 2021

The plate-type fuel assembly adopted in nuclear research reactor suffers from complicated effect induced by non-uniform irradiation, which might affect its stress conditions, mechanical behavior and thermal-hydraulic performance. A reliable numerical method is of great importance to reveal the complex evolution of mechanical deformation, flow redistribution and temperature field for the plate-type fuel assembly under non-uniform irradiation. This paper is the first part of a two-part study developing the numerical methodology for the thermal-fluid-structure coupling behaviors of plate-type fuel assembly under irradiation. In this paper, the thermal-fluid-structure coupling methodology has been developed for plate-type fuel assembly under non-uniform irradiation condition by exchanging thermal-hydraulic and mechanical deformation parameters between Finite Element Model (FEM) software and Computational Fluid Dynamic (CFD) software with Mesh-based parallel Code Coupling Interface (MpCCI), which has been validated with experimental results. Based on the established methodology, the effects of non-uniform irradiation and fluid were discussed, which demonstrated that the maximum mechanical deformation with irradiation was dozens of times larger than that without irradiation and the hydraulic load on fuel plates due to differential pressure played a dominant role in the mechanical deformation.

• Tribology and Lubricants
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• 제25권6호
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• pp.421-426
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• 2009

The brake system is important part of automobile safety system. The disc brake system is divided two parts: the rotating axisymmetrical disc and the stationary pads. During braking, the kinetic energy and potential energy of moving vehicle were converted into the thermal energy through frictional heat between the brake disc and the pads. The frictional heat, which is generated on the interface of the disc and pads, can cause high temperature during the braking process. The object of present work is to determine temperature and thermal stress, to compare to simulation results and experimental results in the disc by partial 3D model of ventilated disc brake with appropriate boundary conditions. In the simulation process, the mechanical loads were applied to the thermo-mechanical coupling analysis in order to simulate the process of heat produced by friction.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.641-646
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• 1998

In order to enhance the dynamic and interactive simulation capability of a system thermal hydraulic code for nuclear power plant, applicability of flow network models in SINDA/FLUIN $T^{™}$ has been tested by modeling feedwater system and coupling to DSNP which is one of a system thermal hydraulic simulation code for a pressurized heavy water reactor. The feedwater system is selected since it is one of the most important balance of plant systems with a potential to greatly affect the behavior of nuclear steam supply system. The flow network model of this feedwater system consists of condenser, condensate pumps, low and high pressure heaters, deaerator, feedwater pumps, and control valves. This complicated flow network is modeled and coupled to DSNP and it is tested for several normal and abnormal transient conditions such turbine load maneuvering, turbine trip, and loss of class IV power. The results show reasonable behavior of the coupled code and also gives a good dynamic and interactive simulation capabilities for the several mild transient conditions. It has been found that coupling system thermal hydraulic code with a flow network code is a proper way of upgrading simulation capability of DSNP to mature nuclear plant analyzer (NPA).

• 한국안전학회지
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• 제9권1호
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• pp.146-154
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• 1994

A series of numerical calculations are performed in order to investigate the dispersion mechanism of toxic gaseous and solid pollutants in extremely short-term and short range. The calculations are carried out in an open space characterized by turbulent boundary layer. The simulation is made by the use of numerical model, in which a control-volume based finite difference method is used together with the SIMPLEC algorithm for the resolution of the pressure-velocity coupling problem. The Reynolds stresses are solved by two-equation, k-$\varepsilon$ model modified for buoyancy. The major parameters consider-ed in this study are temperature, velocity and Injection height of toxic gases, environmental conditions such as temperature and velocity of free stream air, and topographic factor. The results are presented and discussed in detail. The flow field is commonly characterized by the formation of a strong recirculation zone due to the upward motion of the hot toxic gas and ground shear stress. The driving force of the upward motion is explained by the effect of thermal buoyancy of hot gas and the difference of inlet velocity between toxic gas and free stream.

• 터널과지하공간
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• 제34권2호
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• pp.127-142
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• 2024

수치해석 기법의 강건성은 다양한 모델링 조건에서 계산 성능이 유지되는 것을 의미하며, 새로운 해석기법 또는 수치코드는 벤치마크 테스트를 통해 강건성이 평가될 필요가 있다. TOUGH-FLAC 모델링 기법은 국내외적으로 이산화탄소 지중저장, 사용후핵연료 지층처분, 지열 개발 등 다양한 분야에 적용되었으며, 실험 계측자료, 다른 수치코드들과의 결과 비교를 통해 모델링 유효성이 분석되었다. 본 연구에서는 해석해를 갖는 열-수리-역학적 복합거동 문제를 토대로 TOUGH-FLAC 기법의 벤치마크 테스트를 수행하였다. 적용된 해석해는 완전히 포화된 지반에 점열원 작용 시 주변매질의 온도, 간극수압, 역학적 거동과 관계되며, 해석해와 수치모사 결과를 비교하여 TOUGH-FLAC 기법의 강건성이 평가되었다. 또한, 열-수리-역학 해석의 연계항, 유체 상변화, 시간증분이 복합거동 계산에 미치는 영향을 조사하였다.

• Geomechanics and Engineering
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• 제16권4호
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• pp.363-373
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• 2018

The evaluation of Thermo-Hydro-Mechanical (THM) coupling behavior is important for the development of underground space for various purposes. For a high-level radioactive waste repository excavated in a deep underground rock mass, the accurate prediction of the complex THM behavior is essential for the long-term safety and stability assessment. In order to develop reliable THM analysis techniques effectively, an international cooperation project, Development of Coupled models and their Validation against Experiments (DECOVALEX), was carried out. In DECOVALEX-2015 Task B2, the in situ THM experiment that was conducted at Horonobe Underground Research Laboratory(URL) by Japan Atomic Energy Agency (JAEA), was modeled by the research teams from the participating countries. In this study, a THM coupling technique that combined TOUGH2 and FLAC3D was developed and applied to the THM analysis for the in situ experiment, in which rock, buffer, backfill, sand, and heater were installed. With the assistance of an artificial neural network, the boundary conditions for the experiment could be adequately implemented in the modeling. The thermal, hydraulic, and mechanical results from the modeling were compared with the measurements from the in situ THM experiment. The predicted buffer temperature from the THM modelling was about $10^{\circ}C$ higher than measurement near by the overpack. At the other locations far from the overpack, modelling predicted slightly lower temperature than measurement. Even though the magnitude of pressure from the modeling was different from the measurements, the general trends of the variation with time were found to be similar.

• Nuclear Engineering and Technology
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• 제56권3호
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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.