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Conference Article: Analysis of cooling capability of reactor vessel cooling system of the experimental VHTR

Suzuki, K.; Sumimoto, H.; Hirano, M. (Department of Power Reactor Projects, Japan Atomic Energy Research Inst., Tokai, Ibaraki); Asami, T. (Babcock Hitachi K.K., Tokyo (Japan))

Abstract

In the Experimental VHTR being developed in JAERI, a Reactor Vessel Cooling System (RVCS) is provided to remove residual heat and fission product decay heat from the core at the postulated accidents, e.g. rapid depressurization accident or total loss of forced circulation accident. The RVCS is one of the engineered safety features of the VHTR. Cooling capability of the RVCS is analyzed. The primary means of heat transfer from the core is heat conduction of the core structure materials and radiation heat transfer between the reactor vessel wall and the RVCS cooling water tube surface being kept below 60 deg. C. For this analysis, a two dimensional R-Z cylindrical model was used. This paper shows calculational methods for the effective thermal conductivity and the natural convection flow in the core and some analytical results of the total loss of forced circulation accident are also presented.

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key words: after-heat removal; computer calculations; cooling systems; loss of flow; natural convection; pressure vessels; reactor safety; temperature distribution; very high temperature; vhtr reactor; accidents; containers; convection; energy transfer; enriched uranium reactors; experimental reactors; gas cooled reactors; graphite moderated reactors; heat transfer; helium cooled reactors; htgr type reactors; power reactors; reactor accidents; reactors; research and test reactors; safety; thermal reactors
Reference:
Specialists' meeting on safety and accident analysis for gas-cooled reactors. Oak Ridge, TN (USA). 13-15 May 1985
International Atomic Energy Agency, Vienna (Austria)
IAEA-TECDOC--358, pp:251-259