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AbstractAbstract
[en] The thermalhydraulics code ASTEC is used to predict flows for a 2 MW natural circulation test carried out in the prototype fast reactor (PFR). The simulation corresponds to a period 2 hours into the test when quasi-steady state conditions are well established. A finite element mesh, representing a 600 sector of the PFR inner pool together with an associated intermediate heat exchanger, is used for a 3-dimensional calculation where buoyancy forces dictate both flow paths and flow rates. The predicted flow field is in close agreement with a 2-loop mode of flow. The predictions are also in good agreement with the experimental results, and confirm the 2-loop mode explanation for some anomalous temperature measurements. (author)
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Source
Apr 1989; 36 p; 4. International meeting on nuclear reactor thermal hydraulics; Karlsruhe (Germany, F.R.); 10-13 Oct 1989
Record Type
Report
Literature Type
Conference
Report Number
Country of publication
ALKALI METALS, BREEDER REACTORS, COMPUTER CODES, CONVECTION, COOLING SYSTEMS, ELEMENTS, ENERGY TRANSFER, EPITHERMAL REACTORS, FAST REACTORS, FBR TYPE REACTORS, FLUID FLOW, HEAT TRANSFER, INFORMATION, LIQUID METAL COOLED REACTORS, LMFBR TYPE REACTORS, MATHEMATICAL MODELS, METALS, NUMERICAL SOLUTION, POWER REACTORS, REACTOR COMPONENTS, REACTORS, SIMULATION, SODIUM COOLED REACTORS
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