Natural-circulation loops modeling

PiUdiwal, D.S., Ambrosini, W., Forgione, N., Vijayan, P.K., Saha, D., Ferreri, J.C., 2007. Analysis of the unstable behavior of a single-phase natural circulation loop with one-dimensional and computational fluid-dynamic models. Annals of Nuclear Energy 34, 339—355. 

During normal operation of the reactor, core heat is removed by the natural circulation of Pb-Bi coolant. The coolant at 1173 K enters the fuel tube in the lower plenum, absorbs the reactor heat, and at 1273 K reaches the upper plenum. Twelve sodium heat pipes transfer heat from the upper plenum to the system of heat utilizing vessels. Thermal-hydraulic analyses were carried out to study natural circulation and the effect of orificing in the primary loop. A computer model based on the law of conservation of momentum was developed for this analysis a simplified model of the primary loop is shown in Fig. XXIX-7. 

The energy conservation model equations are written for the directions of flow shown in Fig. 16.4. To save on the equation processing, the case of loss and injection of fluid from/to the loops is not considered in detail. The situation as outlined for the mass conservation equation holds by direct analogy. Note that net loss, without replacement, of mass and energy from either loop can lead to situations for which the intended natural circulation will not be present. 

In our assessment [10], MELCOR correctly calculated the thermal/hydraulic phenomena observed during steady-state, single-phase liqiud natural circulation, as summarized in Table 3.1. MELCOR predicted the correct total flow rate and the flow split between two unequal loops without any ad hoc adjustment of the input. The code could reproduce the major ther-mal/hydraulic response characteristics in two-phase natural circulation, after a number of nonstandard input modelling modifications MELCOR could not reproduce the requisite physical phenomena with normal input models. The natural circulation mass flows predicted in these two cases are shown in Figure 3.1. 

The RELAP5 model was used to simulate the thermal-hydraulics of the reactor vessel, the piping in all three primary coolant loops, the pressurizer, all three steam generators, and selected parts of the secondary systems. Reactor vessel nodalization, as developed by Bayless, is shown in Figure 1. As indicated, three parallel flow channels extend from the lower plenum through the core to the upper reactor vessel head. If the appropriate conditions exist, this arrangement will allow development of in-vessel natural circulation. 

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