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A Nodal Kinetics and Thermohydraulics Analysis (NOKTA) Code for Analyzing Rod-Ejection Accidents and Other Transients in Nuclear Power Reactor Cores

Sadi Kaya, Hasbi Yavuz

Nuclear Technology

Volume 129 / Number 1 / January 2000 / Pages 26-35


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For analyzing nuclear power reactor core transients, a three-dimensional nodal kinetics and thermohydraulics code, NOKTA, was developed. Nodal kinetics calculation is based on a one-group neutron diffusion approach. Thermal-hydraulics analysis is handled as in the COBRA-IV-I code. The NOKTA code was designed for analyzing especially large reactivity accidents, such as sudden rod ejection. It can also analyze intermediate transients, such as sharp power changes that may initiate xenon oscillations, and slow transients, such as boric acid density changes in the flow. The code dimensions are set at 125 subchannels and 30 axial levels. Calculation starts with a saturated xenon density, one-group neutronics parameters, and a flux profile, which is required as an input. Initially, keff of each computation cell is set to unity.

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