COBRA-OSU : a fast running computer code for coupled kinetic-thermal hydraulic analysis of nuclear reactor cores Public Deposited

http://ir.library.oregonstate.edu/concern/graduate_thesis_or_dissertations/qz20sv76b

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  • COBRA-OSU computer code is a fast running computer code for coupled Kinetic-Thermal Hydraulic Analysis of nuclear reactor cores. This code which is based on the standard version of COBRA-IV[superscript (3)] has two major improved features. First, the COBRA-OSU computer code uses Gaussian elimination method instead of Gauss-Siedel iteration for crossflow calculations. Second, the code has an additional model for analysis of transient neutronic behavior of the reactor core. This model is a region-wise point reactor kinetics which calculates the total feed-back reactivity of the core at different axial locations and solves for the axial power distribution at each transient time step. COBRA-OSU runs five to thirty times faster and provides results which are more accurate as compared to those of COBRA-IV. This fast numerical convergence in COBRA-OSU is a direct result of a new approach for the calculation of the crossflows for which COBRA-IV uses many external iterations to converge. COBRA-OSU also handles feed-back reactivity effects on the axial power profile during the course of a transient. This transient event can be a rod ejection accident, any kind of line break accident, a turbine or pumps trip, a matter of a change in plant power level and so on. The operating system of a nuclear reactor must be able to properly respond to any kind of transient conditions to be licensed. Feed-back reactivity effect on core power generation during the course of a transient is of primary importance in accurate calculation of reactor status parameters. It is also helpful in design and operation of a nuclear reactor, especially for power maneuvering optimization of large reactor power systems.
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