Thermal-hydraulic aspects of flow inversion in a research reactor
PARET, a neutronics and thermal-hydraulics computer code, has been modified to account for natural convection in a reactor core. The code was then used to analyze the flow inversion that occurs in a reactor with heat removal by forced convection in the downward direction after a pump failure. Typical results are shown for a number of parameters. Research reactors normally operating much above ten MW are predicted to experience nucleate boiling in the event of a flow inversion. Comparison with experimental results from the Belgian BR2 reactor indicated general agreement although nucleate boiling that was analytically predicted was not noted in the BR2 data.
- Publication Date:
- OSTI Identifier:
- Report Number(s):
ON: DE87004751; TRN: 87-008016
- DOE Contract Number:
- Resource Type:
- Resource Relation:
- Conference: Reduced Enrichment for Research and Test Reactors (RERTR) program international meeting, Gatlinburg, TN, USA, 3 Nov 1986
- Research Org:
- Argonne National Lab., IL (USA)
- Country of Publication:
- United States
- 21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; 42 ENGINEERING; FLUID FLOW; P CODES; REACTOR COOLING SYSTEMS; RESEARCH AND TEST REACTORS; COMPUTERIZED SIMULATION; HEAT TRANSFER; HYDRAULICS; COMPUTER CODES; COOLING SYSTEMS; ENERGY SYSTEMS; ENERGY TRANSFER; FLUID MECHANICS; MECHANICS; REACTOR COMPONENTS; REACTORS; SIMULATION 220600* -- Nuclear Reactor Technology-- Research, Test & Experimental Reactors; 420400 -- Engineering-- Heat Transfer & Fluid Flow
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