抄録
An universally applicable thermal-hydraulic 'Coolant Channel Module' and code package CCM has been constructed with the objective to simulate in an as general as possible way the steady state and transient behaviour of characteristic parameters of a single- or two-phase fluid flowing within any type of heated or non-heated coolant channel. It can then be used to describe in a very effective way the thermal-hydraulic situation of a number of complex physical systems. These can be different types of steam generators, 1D and 3D nuclear reactor cores with special attention to the calculation of the mass flow distribution into different parallel channels after non-symmetric perturbations etc. Applied within the natural-circulation U-tube steam generator code UTSG-3 three of the most characteristic channel elements of such an U-tube steam generator, namely the primary and secondary side of the heat exchange region and the riser region, have been represented by adequate CCM modules. The way of working of the code UTSG-3 (and thus the module CCM) will be demonstrated on hand of the characteristic test case 'Loss of main feedwater (but auxiliary and emergency feedwater cooling still available) with turbine trip but a defect in the scram system (ATWS case)' followed by a corresponding sequence in BOP actions.
