Home / Store / Journals / Electronic Articles / Nuclear Technology / Volume 146 / Number 3 / Pages 267-278
Jorge Solís, Maria N. Avramova, Kostadin N. Ivanov
Nuclear Technology / Volume 146 / Number 3 / Pages 267-278
Format:electronic copy (download)
A multilevel methodology has been developed to extend the TRAC-BF1/NEM coupled code capability to obtain the transient fuel rod response. The COBRA-TF thermal-hydraulics subchannel analysis code is coupled to TRAC-BF1/NEM in the parallel virtual machine environment. The power information obtained from the nodal expansion method three-dimensional neutronic calculation is used by the hot subchannel analysis module. The TRAC-BF1 thermal-hydraulic system analysis code provides the COBRA-TF thermal-hydraulic boundary conditions. The subchannel analysis module uses this information to recalculate the fluid, thermal, and hydraulics conditions in the most limiting node (axial region of assembly/channel) within the core at each time step. A dynamic algorithm has been developed to identify the most limiting channel and fuel assembly (radially) and axial region (node) based on the current state of the core. Results, obtained with the new parallel multilevel coupled methodology, are presented and discussed for the Mexican Laguna Verde 1 nuclear power plant control rod drop accident.
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