An extension of the nodal Green's function method for reactor analysis
A method is presented for calculating the nodal flux distribution and the pin power distribution, as well as the effective multiplication, in a nuclear power reactor described by the one-dimensional, two-group diffusion equation. The method is based on the use of Green's functions in a nodal reactor description, and it extends the work of previous authors by including burnup-induced heterogeneities and by calculating local pin power distributions from spatial flux distributions within the node obtained by piecewise polynomial interpolation. An advantage of the method is that one obtains power and exposure distributions at fine mesh points, while retaining the economy characteristic of solutions of the neutron diffusion equation in the nodal framework. In numerical calculations carried out on model problems, good agreement is achieved between the results of the extended nodal Green's function method and those obtained using the CITATION finite difference code.
- Research Organization:
- University of California, Department of Nuclear Engineering Berkeley, California 94720
- OSTI ID:
- 6835239
- Journal Information:
- Nucl. Sci. Eng.; (United States), Vol. 81:2
- Country of Publication:
- United States
- Language:
- English
Similar Records
A Variational Nodal Expansion Method for the Solution of Multigroup Neutron Diffusion Equations with Heterogeneous Nodes
Whole-core comparisons of subelement and fine-mesh variational nodal methods
Related Subjects
REACTOR KINETICS
GREEN FUNCTION
NEUTRON DIFFUSION EQUATION
BURNUP
C CODES
COMPUTER CALCULATIONS
COMPUTER CODES
FINITE DIFFERENCE METHOD
FUEL PINS
HETEROGENEOUS EFFECTS
MESH GENERATION
MULTIGROUP THEORY
MULTIPLICATION FACTORS
NEUTRON FLUX
ONE-DIMENSIONAL CALCULATIONS
POWER DISTRIBUTION
POWER REACTORS
REACTOR KINETICS EQUATIONS
SPATIAL DISTRIBUTION
DIFFERENTIAL EQUATIONS
DISTRIBUTION
EQUATIONS
FUEL ELEMENTS
FUNCTIONS
ITERATIVE METHODS
KINETICS
NEUTRON TRANSPORT THEORY
NUMERICAL SOLUTION
RADIATION FLUX
REACTOR COMPONENTS
REACTORS
TRANSPORT THEORY
220300* - Nuclear Reactor Technology- Fuel Elements
220100 - Nuclear Reactor Technology- Theory & Calculation