Pressure drop characterization of Advanced Fuel for Argentine Reactors

Full text: The Advanced Fuel for Argentine Reactors (CARA) is being developed by National Atomic Energy Commission (CNEA) of Argentina, involving different engineering and development groups. One of the main objectives of the design is to maintain the hydraulic pressure drop per fuel channel of Embalse (PHWR horizontal pressure channel) and Atucha I (PHWR pressure vessel with vertical channel) nuclear power plants, to enable its use in both reactor types. The CARA fuel bundle has 52 fuel rods, and its length is twice that of Embalse. At the extremes it has end plates, while at intermediate positions solid spacers of CANDU design were replaced by elastic spacer PWR type. The Embalse fuels take alertly azimuthal rotation between neighbour fuels inside the refrigerant channel. The end plates and the misalignment between fuels cause discontinuities in flow sub channels. Hence the pressure drop due to this bundle junction is a function of azimuthal misalignment angle. The restrictive pressure drop given by Atucha I is lower than the Embalse one, and the possibility to fix five fuels completely aligned by a coupling system, suggest to achieve lower pressure drop at the junctions, in order to allow the CARA use in both Argentine NPPs. This possibility becomes very important for the hydraulic characterization of the CARA fuel bundle in the first stage of the project. A one dimensional analytical model for local and distributed friction has been developed. It consider specially the junction pressure drop between adjacent fuels bundles as function of misalignment angle. This model allows to predict both cases in question, aleatory and fully aligned for Embalse and Atucha I respectively. This work presents validation results of end plate pressure drop model as a function of azimuthal misalignment and the friction pressure drops model compared with experimental data. Using experimental data measured for Embalse fuel at CNEA low pressure experimental test loop and published data for the CANDU fuel, the results of the model were compared with reference values of our original fuels inside each nuclear power plant coolant channel. The CARA pressure drop was estimated at low pressure loop conditions, and at both channel reactor conditions using the validated model. It demonstrated its compatibility in both cases