**5. Conclusion**

Flow instability is an important consideration in the design of nuclear reactors due to the possibility of flow excursion during postulated accident. In MTR, the safety criteria will be determined for the maximum allowable power and the subsequent analysis will therefore restrict to the calculations of the flow instability margin. In the present work, a new empirical correlation to predict the subcooling at the onset of flow instability in vertical narrow rectangular channels simulating coolant channels of MTR was developed. The developed correlation involves almost all parameters affecting the phenomenon in a dimensionless form and the coefficients involved in the correlation are identified by the experimental data of Whittle and Forgan that covers the wide range of MTR operating conditions. The correlation predictions for subcooling at OSV were compared with predictions of some previous correlations where the present correlation gives much better agreement with the experimental data of Whittle and Forgan with relative standard

Flow Instability in Material Testing Reactors 43

Ahmad, S. Y. (1970). Axial distribution of bulk temperature and void fraction in a heater channel with inlet subcooling, Journal of Heat Transfer, Vol. 92, pp. 595-609. Babelli, I. & Ishii, M. (2001). Flow excursion instability in downward flow systems Part I. Single-phase instability, Nuclear Engineering and Design, Vol. 206 pp. 91-96. Bergisch Gladbach (April 1992). Safety Analyses for the IAEA Generic 10 MW Reactor,

Bowering, R. W. (1962). Physical model based on bubble detachment and calculation of

Chang, H. OH & Chapman, J. C. (1996). Two-Phase Flow Insatiability for Low–Flow Boiling

Chatoorgooon, V.; Dimmick, G. R.; Carver, M. B.; Selander, W. N. & Shoukri, M. (1992).

Dilla, E. M.; Yeoh, G. H. & Tu, J. Y. (2006). Flow instability prediction in low-pressure

Doughherty, T.; Fighetti, C.; McAssey, E.; Reddy, G.; Yang, B.; Chen, K. & Qureshi, Z. (1991). Flow Instability in Vertical Channels, ASME HTD-Vol. 159, pp. 177-186.

Void at Low Pressures", Nuclear Technology, Vol. 98, pp.366-378.

steam voidage in the subcooled region of a heated channel", HPR-10, Institute for

in Vertical Uniformly Heated Thin Rectangular Channels, Nuclear Technology,

Application of Generation and Condensation Models to Predict Subcooled Boiling

subcooled boiling flows using computational fluid dynamics code, ANZIAM

ΔP : pressure drop, Pa

: heat flux W/m2

: density, kg/m3

f : liquid phase,

g : vapor phase, h : heated in : inlet

s : saturation, w : wall.

**7. References** 

: time, s

**Subscripts** 

μ : dynamic viscosity, kg/m s

: surface tension, N/m

w : wall shear stress, N/m2

OFI : onset of flow instability, OSV : onset of significant void,

fg : difference of liquid and vapor,

IAEA-TECDOC-643, Vol. 2, Appendix A.

Atomenergi, Halden, Norway.

Jornal, Vol. 46, pp. C1336-C1352.

Vol. 113, pp.327-337.

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.
