**6. Nomenclature**


#### **Greek Letters**



in : inlet

42 Nuclear Reactors

deviation of only 6.6%. The bubble detachment parameter was also estimated based on the present correlation. The present correlation was then utilized in a model predicting the void fraction and pressure drop in subcooled boiling under low pressure. The pressure drop model predicted the S-curves representing the two-phase instability of Whittle and Forgan with good accuracy. The present correlation was also incorporated in the safety analysis of the IAEA 10 MW MTR generic reactor in order to predict the OFI phenomenon under both fast and slow loss-of-flow transient. The OFI locus for the reactor coolant channels was predicted and plotted against flow velocity, exit temperature and bubble detachment parameter for various heat flux values. It was found that the reactor has vast safety margins for OFI phenomenon under both steady and transient

states.

**6. Nomenclature** 

Cp : specific heat, J/kgºC d : gap thickness, m db : bubble diameter, m dh : heated diameter, m de : hydraulic diameter, m g : acceleration of gravity, m/s2

G : mass flux, kg/ms2 I : enthalpy, J/kg

L : active length, m

P : pressure, Pa

Nu : Nusselt number, = *<sup>e</sup> hd k*

Pe : Peclet number, = RePr

Re : Reynolds number = *G de*

St : Stanton number, = *Nu Pe*

z : distance in axial direction, m

: void fraction, dimensionless

Pr : Prantdel number, =

T : temperature, ºC U : coolant velocity, m/s W : channel width, m x : steam quality

**Greek Letters** 

Ifg : latent heat of vaporization, J/kg k : thermal conductivity, W/mºC

> *Cp k*


#### **7. References**


Flow Instability in Material Testing Reactors 45

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Saha, P.; Ishii, M. & Zuber, N. (1976). An Experimental Investigation of the Thermally

Saha, P. & Zuber, N. (1976). An Analytical Study of the Thermally Induced Two-Phase Flow

Saha, P. & Zuber, N. (1974). Point of net vapor generation and vapor void fraction in

Soon Heung; Yun Il Kim & Won-Pil Beak (1996). Derivation of mechanistic critical heat flux

Sridhar Hari and Yassin A. Hassan (2002). Improvement of the subcooled boiling model for

Staub, F. W. (1968). The Void Fraction in Sub-Cooled Boiling-Prediction of the Initial Point of Net Vapor Generation, J. Heat Transfer, Trans. ASME, Vol. 90, pp. 151-157. Stelling, R.; McAssey, E. V.; Dougherty, T. & Yang, B. W. (1996). The onset of flow instability

Unal, H. C. (1977). Void Fraction and Incipient Point of Boiling During the Subcooled

Vijay Chatoorgoon (1986). SPORTS- A simple non-linear thermal-hydraulic stability code,

Whittle, R. H. & Forgan, R. (1967). A Correlation for the Minima in the Pressure Drop

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**3** 

 *Japan* 

Motoo Fumizawa

*Shonan Institute of Technology* 

**Herium-Air Exchange Flow Rate Measurement** 

Buoyancy-driven exchange flows of helium-air were investigated through horizontal and inclined small openings. Exchange flows may occur following a window opening as ventilation, fire in the room, over the escalator in the underground shopping center as well as a pipe rupture accident in a modular high temperature gas-cooled nuclear reactor. Fuel loading pipe is located in the inclined position in the pebble bed reactor such as Modular

In safety studies of High Temperature Gas-Cooled Reactor (HTGR), a failure of a standpipe at the top of the reactor vessel or a fuel loading pipe may be one of the most critical designbase accidents. Once the pipe rupture accident occurs, helium blows up through the breach immediately. After the pressure between the inside and outside of he pressure vessel has balanced, helium flows upward and air flows downward through he breach into the pressure vessel. This means that buoyancy-driven exchange flow occurs through the breach, caused by density difference of the gases in the unstably stratified field. Since an air stream corrodes graphite structures in the reactor, it is important to evaluate and reduce the air

Some studies have been performed so far on the exchange flow of two fluids with different densities through vertical and inclined short tubes. Epstein(Epstein, 1988) experimentally and theoretically studied the exchange flow of water and brine through the various vertical tubes. Mercer et al. (Mercer, 1975) experimentally studied an exchange flow through inclined tubes with water and brine. He performed the experiment in the range of 3.5 <L/D < 18 and 0 deg < θ < 90 deg, and pointed out that the length-to-diameter ratio L/D, and the inclination angle θof the tube are the important parameter for the exchange flow rate. Most of these studies were performed on the exchange flow with a relatively small difference of the densities of the two fluids (up to 10 per cent). However, in the case of HTGR standpipe rupture accident, the density of the outside gas is at least three times larger than that of the gas inside the pressure vessel. Few studies have been performed so far in such a large range of density difference. Kang et al. (Kang, 1992) studied experimentally the exchange flow through a round tube with a partition plate. Although we may think that the partitioned plate, a kind of obstacle in the tube, decrease the exchange flow rate, he found that the exchange flow rate is increased by the partition plate because of separation of an upward

reactor (Fumizawa, 2005, Kiso, 1999) and AVR(El-Wakil, 1982, Juni-1965, 1965).

ingress flow rate during the standpipe rupture accident.

**1. Introduction** 

and downward flow.

**Through a Narrow Flow Path** 

Zuber, N.; Staub, F. W. & Bijwaard, G. (1966). Vapour void fraction in subcooled boiling systems, Proceeding of the third International heat transfer conference, Vol. 5, pp. 24-38, Chicago.
