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Missouri University of Science and Technology
<p>Boiling water reactors<br />Nuclear fuel rods -- Testing<br />Heat flux -- Research</p>
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Possible Damage Mechanism In Dresden 2 Class Boiling Water Reactor Fuel, Donald Lee Moffett
Possible Damage Mechanism In Dresden 2 Class Boiling Water Reactor Fuel, Donald Lee Moffett
Masters Theses
"A thermal hydraulic analysis of a typical hot channel of a Dresden 2 class boiling water reactor is studied for possible severe local overheating. The COBRA-II thermal hydraulic analysis code is modified to include critical heat flux calculations for each subchannel. The effects of flow distribution, inlet mass velocity variations, dimensional tolerances, enrichment variations, and input parameters are examined in detail. Bulk channel results are in good agreement with the published data, but the assembly wall side of the corner fuel rod has a minimum critical heat flux ratio of less than unity for a number of the situations examined. …