Heat transfer study under supercritical pressure conditions for single rod test section

H. Mori, S. Yoshida, Shinichi Morooka, H. Komita

Research output: Chapter in Book/Report/Conference proceedingConference contribution

10 Citations (Scopus)

Abstract

The supercritical pressure -water cooled power reactor (SCPR), which is a once-through type water cooled reactor supplying high temperature supercritical-pressure water vapor to the turbine system, has an innovative and economical potential. Heat transfer performance at a supercritical pressure is one of the important design items. Many heat transfer studies have been done for fully developed supercritical pressure (SCP) fossil power plant boilers. However, the SCPR conditions are different from those for the SCP fossil boilers in the following three points. 1) The thermal-hydraulic diameter in the SCPR flow channel is smaller than that of the SCP boiler tubes. 2) Coolant flows outside of the fuel rods (outside-flow) for the SCPR while coolant flows inside of the tubes (inside-flow) for the SCP boiler. 3) For the SCPR the heat transfer is influenced from turbulence created by the bundle geometry unlike the SCP fossil boiler. In relation to the development of the SCPR, the authors are carrying out the following experimental studies on the heat transfer to supercritical pressure fluids using a supercritical pressure HCFC22 facility at Kyushu University, Japan: 1) small heater tube inside-flow tests for the influence of thermal-hydraulic diameter,.2) single heater rod tests simulating a fuel rod for the influence of outside-flow and 3) three heater rods tests simulating the SCPR fuel-bundle for the influence of turbulence created by the bundle geometry. Results of the small heater tube tests have already been reported. In this study, to obtain the data using single heater rod simulating a fuel rod and to evaluate the influence of outside-flow, single heater rod tests were performed using a single rod test section. The heater rod of 8.0 mm outside diameter and 1.8 m heated length was installed in an unheated outer tube of 10 mm inside diameter. The flow section was annulas and its thermal equivalent diameter simulated that of the typical SCPR fuel subchannel. The following results were obtained. (1) For the single rod, the occurrence and degree of the deterioration in heat transfer was suppressed. Therefore, whereas the heat transfer in the 'normal ' region was similar to that of the tube, the heat transfer in the 'deteriorated' region was kept better. (2) The frictional pressure drop of the single rod was not affected by the heat flux, unlike the tube, and can be evaluated by the Itaya correlation for isothermal singe phase flow.

Original languageEnglish
Title of host publicationProceedings of the American Nuclear Society - International Congress on Advances in Nuclear Power Plants 2005, ICAPP'05
Pages1548-1556
Number of pages9
Volume3
Publication statusPublished - 2005
Externally publishedYes
EventAmerican Nuclear Society - International Congress on Advances in Nuclear Power Plants 2005, ICAPP'05 - Seoul
Duration: 2005 May 152005 May 19

Other

OtherAmerican Nuclear Society - International Congress on Advances in Nuclear Power Plants 2005, ICAPP'05
CitySeoul
Period05/5/1505/5/19

ASJC Scopus subject areas

  • Nuclear Energy and Engineering
  • Safety, Risk, Reliability and Quality

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  • Cite this

    Mori, H., Yoshida, S., Morooka, S., & Komita, H. (2005). Heat transfer study under supercritical pressure conditions for single rod test section. In Proceedings of the American Nuclear Society - International Congress on Advances in Nuclear Power Plants 2005, ICAPP'05 (Vol. 3, pp. 1548-1556)