Development of thermohydraulics computer programs for thermal striping phenomena
- Power Reactor and Nuclear Fuel Development Corp., O-arai, Ibaraki (Japan). O-arai Engineering Center
- Tokyo Inst. of Tech., Tokyo (Japan). Research Lab. for Nuclear Reactors
Thermal striping phenomena characterized by stationary random temperature fluctuations are observed in the region immediately above the core exit of liquid-metal-cooled fast reactors (LMFRs) due to the interactions of cold sodium flowing out of a control rod (C/R) assembly and hot sodium flowing out of adjacent fuel assemblies (F/As). Two thermohydraulics computer programs AQUA and DINUS-3, which are represented by both time- and volume-averaged transport analysis and direct numerical simulation of turbulence, respectively, were developed and validated for the evaluation of thermal striping phenomena. These codes were incorporated with higher order difference schemes to approximate the convection terms in conservation equations and adaptive time-step size control systems based on the fuzzy theory to eliminate numerical instabilities. From validation analyses with fundamental experiments in water and sodium, it was concluded that (a) thermal striping conditions such as spatial distributions of the intensity and the frequency of the fluid temperature fluctuations can be estimated efficiently by a combined approach incorporating the AQUA code and the DINUS-3 code, and (b) the thermal striping phenomena for the in-vessel components of actual liquid-metal-cooled fast reactors can be evaluated by the numerical method without conventional approaches such as large scale model experiments using sodium.
- OSTI ID:
- 201348
- Journal Information:
- Nuclear Technology, Vol. 113, Issue 1; Other Information: PBD: Jan 1996
- Country of Publication:
- United States
- Language:
- English
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