The subchannel analysis method is an important tool for predicting the thermal hydraulic performance of rod bundle nuclear fuel element. It considers a rod bundle to be a continuously interconnected set of parallel flow subchannels which are assumed to contain one dimensional flow and are coupled to each other by crossflow mixing; the axial length is divided into a number of increments such that the whole flow space of a rod bundle is divided into a number of nodes.

The principle of subchannel analysis is the application of continuity and conservation equations to the flow between these nodes. The conservation equations relate the local variations of velocity and enthalpy of each node to those of its neighboring ones. The relation between subchannel flow rate which is the mass flow rate in an axial direction through subchannel area and diversion cross-flow which is the mass flow in a transverse direction resulting from local pressure differences between two subchannels, is strongly governed by momentum balance in a transverse direction.

A correct formulation of momentum equations and good knowledge of the mixing process between subchannels are an absolute necessity, for obtaining reliable predictions using subchannel calculations.

#### REFERENCES

Courtaud, M., Ricque R., Martinet, B. (1966) Etude des pertes de charge dans ies conduites circulates contenant un faisceau de barreaux, *Chem. Eng. Sci.*, 21, 881–893. DOI: 10.1016/0009-2509(66)85082-0

Giot, M. (1981) Friction factors in single channels, *Thermohydraulics of Two-Phase Systems for Industrial Design and Nuclear Engineering*, Delhaye, J. M. et al., Eds., Hemisphere.

Grand, D. (1981) Pressure drops in rod bundles, *Thermohydraulics of Two-Phase Systems for Industrial Design and Nuclear Engineering*, Delhaye, J. M et al., Eds., Hemisphere.

Lafay, J. (1974) Influence de la variation de la viscosite avec la temperature sur le frottement avec transfert de chaleur en regime turbulent etabli, *Int. J. Heat and Mass Trans.*, 17, 815–834. DOI: 10.1016/0017-9310(74)90150-1

Lafay, J., Menant, B. and Barroil, J. (1975) Local pressure measurements and peripheral flow visualization in a water 19 rod bundle compared with FLICA IIB calculations: influence of helical wire wrap spacer system, *ASME Heat Transfer Conf., San Francisco* 75-HT-22.

Reddy, D. G. and Fighetti, C. R. (1983) Evaluation of two-phase pressure drop correlations for high pressure steam-water systems, *ASME Thermal Engineering Conf. Proc, Honolulu*, Vol. 1.

Rheme, K. (1973) Pressure drop correlations for fuel element spacers, *Nuclear Technology*, 17, 15–23, January.

Rogers, J. T. and Todreas, N. E. (1968) Coolant interchannel mixing in reactor fuel rod bundles single phase coolants,* Winter Annual Meeting of the ASME, New York*, December.

Rouhani, Z. (1973) A review of momentum balance in subchannel geometry, *European Two-Phase Flow Group Meeting Brussels*, June 4–7.

Rowe, D. S. (1973) Measurement of turbulent velocity, intensity and scale in rod bundle flow channels, *BNWL-1736*, May.

Tong, L. S., (1968) Pressure drop performance of a rod bundle, *Winter Annual Meeting of the ASME, New-York*, December.