9.135. In the preceding sections, the thermal-hydraulic behavior of a coolant channel associated with a single fuel rod was considered. For re­actor core analysis, however, a more detailed treatment is required. For this purpose, subchannel analysis is commonly used. A subchannel may consist of the coolant flow “cell” associated with a single rod as considered in Example 9.5, or a cluster of such cells. A number of parallel interacting subchannels running the length of the fuel rods may be included in the treatment. In addition, each subchannel length is divided into a number of axial intervals or “control volumes.” The result of this subdivision is a two-dimensional network of thermally and hydraulically connected nodes (§9.59) in the axial and radial directions. Heat transfer occurs primarily by conduction and convection in the radial direction from fuel to coolant. Allowance is also made for cross flow and turbulent mixing between coolant subchannels.

9.136. Numerous subchannel design codes have been written for solving the thermal-hydraulic equations based on the conservation of mass, energy, and momentum for each nodal volume [19]. Solutions yield such quantities as the radial and axial variations in the fluid enthalpy and mass velocity.

The approach to burnout conditions can then be determined by means of an appropriate critical heat flux correlation.

9.137. For safety analysis purposes, it is important to couple the core with the remainder of the coolant system and to consider transient con­ditions. System codes, which extend the subchannel approach, have been developed to meet this need. Typical such codes are RETRAN [20], which was developed under Electric Power Research Institute sponsorship and RELAP [21]. The development effort has been supported by an experi­mental program designed to simulate accident conditions, and the use of operating plant data for verification. Further consideration of system codes for safety analysis will be given in Chapter 12.