Computer codes COBRA

The analysis of transient flows is necessary for safety analysis of nuclear reactors. Such efforts usually result in the development of large computer codes (e.g., RELAP-5, RETRAN, COBRA, TRAC). Rather than going into the details of such codes, this section gives the principles and basic models involved in the analysis. 

Bowring, R., and P. Moreno, 1976, COBRA-IIIC/MIT Computer Code Manual, MIT Department of Nuclear Engineering, Cambridge, MA. (5)... 

Plant safety assessment projects improve the abilities of designers, operators, and regulators to evaluate the safety of their plants through the use of internationally accepted methodologies and computer analysis codes. An example of a key accomplishment in this area is training technical personnel to use the COBRA-SFS computer code to perform thermal-hydraulic analysis (shown in Fig. 4). 

For both submerged and emerged LCBs, accurate predictions with high computational efforts can be obtained from the 2DV RANS-VOF code COBRAS by Losada et al. which allows to represent the structure porosity. 

The second phase is in progress currently, which makes full account of the results from the first phase. Computational models were developed and validated on experimental results for hydraulic pressure drop. Critical heat flux margins were calculated by COBRA-IV code and validated with the use of published data on critical power measurements for very different cluster geometries relevant for CARA. Computational probabilistic analysis of fuel performance was performed with the use of BACO code [4,10] and facilitated the determination of dimensional tolerances. 

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