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Experimental Study on Critical Heat Flux of In-Vessel Retention for Large Nuclear Reactor

[+] Author Affiliations
Canhui Sun, Wei Chen, Jun Geng

State Power Investment Corporation Research Institute, Beijing, China

Paper No. ICONE25-67066, pp. V006T08A073; 7 pages
doi:10.1115/ICONE25-67066
From:
  • 2017 25th International Conference on Nuclear Engineering
  • Volume 6: Thermal-Hydraulics
  • Shanghai, China, July 2–6, 2017
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-5784-7
  • Copyright © 2017 by ASME

abstract

During a rare accident involving severe core damage in a nuclear power plant, if the molten core material can be contained within the boundary of the reactor vessel, the severity of the accident is expected to be greatly reduced. Therefore, the severe accident management strategy based on in-vessel retention (IVR) of molten core debris is highly desirable, and has been adopted by advanced reactor designs such as AP1000, APR1400 and CAP1400.

Previous studies assumed a fully developed molten pool, where molten metal and molten oxides separated, and liquid metal form a layer on the top of the oxidic pool. The limiting failure mechanism of the lower head is related to occurrence of boiling crisis on the outer surface of the vessel. The boiling crisis would occur, when the local heat flux through the vessel wall exceeds the critical heat flux value. The evaluation of thermal regime is then done by comparing the heat flux distribution from the molten corium pool and the CHF variation on the outer wall in order to define the available margins to CHF. When a larger power nuclear reactor is designed, the margin of the IVR could be decreased. Therefore, the study on critical heat flux enhancement for lager nuclear reactor is necessary.

In this study, an experimental facility is conducted to analysis the critical heat flux mechanism and enhancement for lager power nuclear reactor. The facility includes two loops, one is primary loop, and the other is auxiliary loop which could import injection flow in the primary loop. The test section is an inclined downward heated rectangular channel, and the angle is variable. A series of critical heat flux experiments are conducted in this facility. The results are compared with previous experiment results and calculation correlation in the large nuclear reactor operating conditions.

Copyright © 2017 by ASME

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