Abstract
In this article, two-dimensional partial differential equations with time representation of nuclear power reactor kinetics are considered for spatial reactor dynamics and thermo hydraulic behavior analysis of a large thermal advanced gas cooled reactor (AGR) type used for nuclear power generation. The equations include the neutron flux equation and delayed neutron precursor concentration, together with taking into account the equations to represent the thermo hydraulic behavior of the fuel, coolant and moderator temperatures. These equations are solved numerically using the finite difference method. For time propagation, an implicit method is applied. The desired initial condition for the reactor to stay at stable critical condition is established by finding the correct value of reactivity. The reactivity disturbance effect in the reactor is studied for different cases and presented for high reactivity values. The model was developed for the analysis of a large AGR with 2000 MWe for future power generation. The results show that the model not only behaves stably but also predicts the results physically for all the various parameters.
Highlights
Nuclear power stations play an important role in electricity generation in industrial countries
Two-dimensional partial differential equations with time representation of nuclear power reactor kinetics are considered for spatial reactor dynamics and thermo hydraulic behavior analysis of a large thermal advanced gas cooled reactor (AGR) type used for nuclear power generation
Comparison of results obtained by RELAP5 modeling and experimental data collected by Groudev [3] for pressurized water reactors (PWR) of VVER440/V230 type showed that the analysis using this computer code is valid
Summary
Nuclear power stations play an important role in electricity generation in industrial countries. There are some computer codes that have been developed by universities or research establishments in order to predict flow regimes, hydro dynamical instability, reactor kinetics, etc These computer codes (analyzers) include MINCS, PHOENICS, TRAC, RELAP5, and others, and they predict the optimized criteria for the thermo hydraulic condition for nuclear reactors and fill in the gaps between the scarce experimental results that have been found by different research sources and obtained under substantially different conditions. Comparison of results obtained by RELAP5 modeling and experimental data collected by Groudev [3] for pressurized water reactors (PWR) of VVER440/V230 type showed that the analysis using this computer code is valid Because such codes are not accessible, especially for developing countries, modeling of the nuclear reactors with more accurate mathematics and higher physics is required to predict more precise simulation results. The present model will consider symmetric and anti-symmetric reactivity inlet disturbances and their effect on nuclear reactor behaviors
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