Thermal-hydraulics in nuclear power technology presented at 20th National Heat transfer Conference, Milwaukee, Wisconsin, August 2-5, 1981 by National Heat Transfer Conference (20th 1981 Milwaukee, Wis.)

Cover of: Thermal-hydraulics in nuclear power technology | National Heat Transfer Conference (20th 1981 Milwaukee, Wis.)

Published by American Society of Mechanical Engineers in New York, N.Y. (345 E. 47th St., New York 10017) .

Written in English

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Subjects:

  • Nuclear reactors -- Fluid dynamics -- Congresses.,
  • Heat -- Transmission -- Congresses.

Edition Notes

Book details

Statementsponsored by the Heat Transfer Division, ASME ; edited by K.H. Sun ... [et al.].
ContributionsSun, K. H., American Society of Mechanical Engineers. Heat Transfer Division.
Classifications
LC ClassificationsTK9202 .N29 1981
The Physical Object
Paginationv, 86 p. :
Number of Pages86
ID Numbers
Open LibraryOL3788356M
LC Control Number81065616

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The origins of nuclear thermal-hydraulics and its relevance within the technology for the production of electrical power are discussed together with the link with other disciplines.

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Junchong Yu is a Senior Engineer and a member of the Chinese Academy of Engineering. He received his bachelor’s degree from Southeast University inwith majors in Reactor Thermal Hydraulics and Safety Analysis, as well as in Overall Nuclear Power Devices. The final part of the book reviews the application of thermal hydraulics in nuclear technology, with chapters on V&V and uncertainty in SYS TH codes, the BEPU approach, and applications to new reactor design, plant lifetime extension, and accident analysis.

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A monograph presenting an up-to-date overview of the thermal-hydraulic technology that underlies the design, operation, and safety assessment of boiling water nuclear reactors (BWRs), which represent a large fraction of the world's installed nuclear power : Richard T Lahey.

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Heat Transfer Division.;]. The Thermal-Hydraulics of a Boiling Water Nuclear Reactor. 2nd ed. La Grange Park, IL: American Nuclear Society, Readings by Session Readings from the Collier and Thome text are abbreviated "C&T." Readings from the Todreas and Kazimi text are abbreviated "T&K.".

This course covers the thermo-fluid dynamic phenomena and analysis methods for conventional and nuclear power stations. Specific topics include: kinematics and dynamics of two-phase flows; steam separation; boiling, instabilities, and critical conditions; single-channel transient analysis; multiple channels connected at plena; loop analysis including single and two-phase natural.

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Book Description. Thermal Hydraulics of Water-Cooled Nuclear Reactors reviews flow and heat transfer phenomena in nuclear systems and examines the critical contribution of this analysis to nuclear technology development. With a strong focus on system thermal hydraulics (SYS TH), the book provides a detailed, yet approachable, presentation of.

detailed analysis of selected nuclear power plant components. The table below shows the workload in a classroom envisaged in the course “Thermal-Hydraulics in Nuclear Energy Systems” given at the Royal Institute of Technology.

In total, the course covers 24 hours of lectures and 24 hours of exercises performed with. Journals & Books; Help The 15th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH) () Google Scholar.

Frogheri et al., b. IAEA Nuclear Power Technology Development Section Division of Nuclear Power, Department of Nuclear Energy (). Nuclear Energy and Technology (NUCET) provides an international medium for the communication of original research, ideas and developments in all areas of the field of nuclear energy, education and training, science and technology, modelling and benchmarking of nuclear codes.

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UNM's Nuclear Engineering department and Institute for Space and Nuclear Power Studies makes a strong presence at national meeting. Two graduate students, Andrew Hahn and Ragai Altamimi, Research Assistant Professor Timothy Schriener, and Distinguished and Regents' Professor gave talks on the results of ongoing research on the development of a nuclear power .In power producing systems, the most common example of two phase flow is water and steam.

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These areas of study are .

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