Solving Neutron Transport Equation in the Reactor using the Intergral Average Derivative Method
Journal: International Journal of Science and Research (IJSR) (Vol.5, No. 3)Publication Date: 2016-03-05
Authors : Phan Huy Thien;
Page : 678-681
Keywords : The Neutron field in Reactor; Boltzman transport equation;
- Solving Neutron Transport Equation in the Reactor using the Intergral Average Derivative Method
- EXPONENTIAL-TIME DIFFERENCING METHOD FOR SOLVING REACTOR POINT KINETICS EQUATIONS WITH TEMPERATURE AND POWER FEEDBACK AND ITS VALIDATION BY ANALYSINGTHERMAL REACTOR BENCHMARKS
- Diffusion Approximation to Neutron Transport Equation with First Kind of Chebyshev Polynomials
- Diffusion Approximation to Neutron Transport Equation with First Kind of Chebyshev Polynomials
- Second Type Chebyshev Polynomial Approximation to Linearly Anisotropic Neutron Transport Equation in Slab Geometry
Abstract
Space and time dependent neutron diffusion equations with many energy groups and taking into account delayed neutrons are nonlinear partial differential equations. IAD method and finite difference method for simplified equations to partial differential equations often, then be rewritten in matrix form. General solution of differential equations contain the exponential matrix of matrix coefficients. Specific methods applied IAD (average method integral derivative) as an example to determine the flux reactor radius spherical R reflector layer. R-R'
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