ABSTRACT
The multigroup delayed neutron point kinetic equations in the presence of temperature feedback reactivity are a system of stiff nonlinear ordinary differential equations. The neutron flux and the delayed neutron precursor concentration are important parameters for the study of safety and transient behavior of the reactor power. The point kinetic equations of multigroup delayed neutrons with temperature feedback reactivity describe the neutron density representing the reactor power level, time-dependent reactivity, precursor concentrations of multigroup delayed neutrons, and thermodynamic variables that enter into the reactivity equation. The solution of this system of equation is useful for providing an estimation for transient behavior of a reactor power and other systems of variables of the reactor cores that are fairly tightly coupled.
