• Title/Summary/Keyword: Nuclear structure

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Application of the Leak Before Break(LBB) Concept to a Heat Exchanger in a Nuclear Power Plant

  • Kwon, Jae-Do;Lee, Choon-Yeol;Lee, Yong-Son;Sul, Il-Chan
    • Journal of Mechanical Science and Technology
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    • v.15 no.1
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    • pp.10-20
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    • 2001

  • The leak before break(LBB) concept is difficult to apply to a structure with a thin tube that is immersed in a water environment. A heat exchanger in a nuclear power plant is such a structure. The present paper addresses an application of the LBB concept to a heat exchanger in a nuclear power plant. The minimum leaked coolant amount(approximately 37.9 liters) containing the radioactive material which can activate the radiation detector device installed in near the heat exchanger is assumed. A postulated initial flaw size that can not grow to a critical flaw size within the time period to activate the radiation detector is justified. In this case, the radiation detector can activate the warning signal caused by coolant leakage from initially postulated flaws of the heat exchanger. The nuclear plant can safely shutdown when this occurs. Since the postulated initial flaw size can not grow to the critical flaw size, the structural integrity of the heat exchanger is not impeded. Particularly the informational scenario presented in this paper discusses an actual nuclear plant.

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INTEGRITY ANALYSIS OF AN UPPER GUIDE STRUCTURE FLANGE

  • LEE, KI-HYOUNG;KANG, SUNG-SIK;JHUNG, MYUNG JO
    • Nuclear Engineering and Technology
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    • v.47 no.6
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    • pp.766-775
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    • 2015

  • The integrity assessment of reactor vessel internals should be conducted in the design process to secure the safety of nuclear power plants. Various loads such as self-weight, seismic load, flow-induced load, and preload are applied to the internals. Therefore, the American Society of Mechanical Engineers (ASME) Code, Section III, defines the stress limit for reactor vessel internals. The present study focused on structural response analyses of the upper guide structure upper flange. The distributions of the stress intensity in the flange body were analyzed under various design load cases during normal operation. The allowable stress intensities along the expected sections of stress concentration were derived from the results of the finite element analysis for evaluating the structural integrity of the flange design. Furthermore, seismic analyses of the upper flange were performed to identify dynamic behavior with respect to the seismic and impact input. The mode superposition and full transient methods were used to perform time-history analyses, and the displacement at the lower end of the flange was obtained. The effect of the damping ratio on the response of the flange was also evaluated, and the acceleration was obtained. The results of elastic and seismic analyses in this study will be used as basic information to judge whether a flange design meets the acceptance criteria.

  • https://doi.org/10.1016/j.net.2015.06.012 인용 KSCI

Multi-scale simulation of wall film condensation in the presence of non-condensable gases using heat structure-coupled CFD and system analysis codes

  • Lee, Chang Won;Yoo, Jin-Seong;Cho, Hyoung Kyu
    • Nuclear Engineering and Technology
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    • v.53 no.8
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    • pp.2488-2498
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    • 2021

  • The wall film-wise condensation plays an important role in the heat transfer processes of heat exchangers, refrigerators, and air conditioner. In the field of nuclear engineering, steam condensation is often utilized in safety systems to remove the core decay heat under both transient and accident conditions. In particular, passive containment cooling system (PCCS), are designed to ensure containment safety under severe accident conditions. A computational fluid dynamics (CFD) scale analysis has been conducted to calculate the heat transfer rate of the PCCS. However, despite the increase in computing power, there are challenges in the long-term transient simulation of containment using CFD scale codes. In this study, a heat structure coupling between the CFD and system analysis codes was performed to efficiently analyze PCCS. In addition, the component unstructured program for interfacial dynamics (CUPID) was improved to analyze the condensation behavior of ternary gas mixtures. Thereafter, the condensation heat transfer on the primary side was calculated using the improved CUPID and CFD code, whereas that on the secondary side was simulated using MARS. Both the coupled codes were validated against the CONAN facility database. Finally, conjugate heat transfer simulations with wall condensation in the presence of non-condensable gases were appropriately performed.

  • https://doi.org/10.1016/j.net.2021.03.001 인용 PDF KSCI

DESIGN STUDY OF AN IHX SUPPORT STRUCTURE FOR A POOL-TYPE SODIUM-COOLED FAST REACTOR

  • Park, Chang-Gyu;Kim, Jong-Bum;Lee, Jae-Han
    • Nuclear Engineering and Technology
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    • v.41 no.10
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    • pp.1323-1332
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    • 2009

  • The IHX (Intermediate Heat eXchanger) for a pool-type SFR (Sodium-cooled Fast Reactor) system transfers heat from the primary high temperature sodium to the intermediate cold temperature sodium. The upper structure of the IHX is a coaxial structure designed to form a flow path for both the secondary high temperature and low temperature sodium. The coaxial structure of the IHX consists of a central downcomer and riser for the incoming and outgoing intermediate sodium, respectively. The IHX of a pool-type SFR is supported at the upper surface of the reactor head with an IHX support structure that connects the IHX riser cylinder to the reactor head. The reactor head is generally maintained at the low temperature regime, but the riser cylinder is exposed in the elevated temperature region. The resultant complicated temperature distribution of the co-axial structure including the IHX support structure may induce a severe thermal stress distribution. In this study, the structural feasibility of the current upper support structure concept is investigated through a preliminary stress analysis and an alternative design concept to accommodate the IHTS (Intermediate Heat Transport System) piping expansion loads and severe thermal stress is proposed. Through the structural analysis it is found that the alternative design concept is effective in reducing the thermal stress and acquiring structural integrity.

  • https://doi.org/10.5516/NET.2009.41.10.1323 인용 PDF KSCI