Investigation of mid-loop operation with loss of RHR at INER integral system test (IIST) facility

Lee, C.H.; Liu, T.J.; Way, Yuan-Shun; Hsia, D.Y.

doi:10.1016/0029-5493(95)01152-8

Cited by 21 publications

(5 citation statements)

References 4 publications

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“…We correlate the degradation factor in terms of the condensate Reynolds number, the air mass fraction, and the Jacob number as given by Eq. (11). Most of our data agree with the predicted values with a RMS error of 27.4%.…”

Section: Discussionsupporting

confidence: 89%

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Local heat transfer during reflux condensation mode in a U-tube with and without noncondensible gases

Lee

Chu

et al. 2006

International Journal of Heat and Mass Transfer

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Section: Discussionsupporting

confidence: 89%

“…More recently, studies in this field have been reported by Pilon et al [5], Chou and Chen [6], Moon et al [7], Thumm et al [8], and Vierow et al [9]. Several larger scale experiments have also been carried out with the aid of integral test facilities such as BETHSY [10] and IIST [11,12].…”

Section: Introductionmentioning

confidence: 94%

Local heat transfer during reflux condensation mode in a U-tube with and without noncondensible gases

Lee

Chu

et al. 2006

International Journal of Heat and Mass Transfer

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“…Moreover, when a NPP undergoes maintenance and refuelling, it is common for the coolant system to be partially drained, which is referred to as a "midloop operation." In this state, the water level of the reactor cooling system (RCS) is above the core but below the elevation of the hot leg [87]. During routine shutdowns and refuelling of a NPP, the residual heat removal system (RHRS) is an active cooling system used to eliminate the decay heat from the core.…”

Section: Condensation In Npps and Relevant Studiesmentioning

confidence: 99%

Steam Condensation Heat Transfer in the Presence of Noncondensable Gases (NCGs) in Nuclear Power Plants (NPPs): A Comprehensive Review of Fundamentals, Current Status, and Prospects for Future Research

Albdour,

Addad,

Alyammahi

et al. 2024

International Journal of Energy Research

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Efficient steam condensation is crucial for safe nuclear power plant (NPP) operations, preventing pressure buildup, overheating, and the release of radioactive materials. However, the presence of noncondensable gases (NCGs), such as air, nitrogen, hydrogen, or helium, can hinder the condensation process by creating a thermal resistance layer that impedes steam diffusion and condensation on the system’s surface. Maximizing the efficiency of steam condensation requires a thorough grasp of the fundamental processes, theories, advancements, and technical hurdles. Therefore, this work thoroughly addresses these needs, with a particular emphasis on addressing the challenges posed by NCGs by dividing the work into four thematic areas. The first theme relates to a comprehensive examination of pure steam condensation phenomena, which includes an exploration of familiar condensation scenarios and various film condensation types. The second theme examines condensation in the presence of NCGs, their mixture properties, and related theories and modelling of heat and mass transfer. The third theme investigates condensation in NPP by exploring passive cooling systems and condensation phenomena under both natural and forced convection conditions during nuclear accidents, the origin of NCGs in NPP and their transportation aspects. This is followed by experimental work related to condensation scenarios and scale. Finally, the last theme looks upon the recent advancements in computational fluid dynamics (CFD) modelling of wall condensation, system analysis codes coupling with CFD, and the implementation of machine learning (ML) for predicting the condensation HTC. By bridging the gap between fundamental knowledge and practical applications, the four thematic areas presented in this work are aimed at providing a comprehensive foundation for researchers and experts in the field of steam condensation when NCGs are involved. The ultimate objective is to bolster the safety and efficacy of NPP operations by understanding the heat and mass transfer mechanisms while mitigating the risk of catastrophic events.

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“…A reduced-height and reduced-pressure IIST facility has been established for safety studies of the Westinghouse three loops pressurized water reactor (PWR) since 1992. The research purposes of the IIST facility are enhancement of understanding of thermal hydraulics phenomena during the accidents [1][2][3], contributing to evaluate and develop the safety computer codes [4,5], and validation of EOP during the accidents of PWR [6]. The IIST facility has three loops as well as all the systems associated with Westinghouse PWR plant system transients.…”

Section: Introductionmentioning

confidence: 99%

Assessment of TRACE CCFL Model with SBLOCA Experiment of IIST Facility

Yang

Wang

Lin

et al. 2012

Science and Technology of Nuclear Installations

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In this paper, the TRACE model for IIST facility is developed and verified with the Small Break loss of coolant accident (SBLOCA) experiment of IIST (Institute of Nuclear Energy Research Integral System Test) facility. By using the Wallis and Kutateladze correlations of countercurrent flow limitation (CCFL) model, the TRACE analyses results, such as break flow rate, primary pressure, and the temperature of cold-leg and hot-leg, are consistent with the IIST data. The results show the Kutateladze correlation of CCFL model can well predict the SBLOCA behavior and present good agreement with IIST experiment data in this paper. Besides, the sensitivity study results of Kutateladze correlation in CCFL model are verified and compared with the IIST data.

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Investigation of mid-loop operation with loss of RHR at INER integral system test (IIST) facility

Cited by 21 publications

References 4 publications

Local heat transfer during reflux condensation mode in a U-tube with and without noncondensible gases

Local heat transfer during reflux condensation mode in a U-tube with and without noncondensible gases

Steam Condensation Heat Transfer in the Presence of Noncondensable Gases (NCGs) in Nuclear Power Plants (NPPs): A Comprehensive Review of Fundamentals, Current Status, and Prospects for Future Research

Assessment of TRACE CCFL Model with SBLOCA Experiment of IIST Facility

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