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Title: DIF3D nodal neutronics option for two- and three-dimensional diffusion theory calculations in hexagonal geometry. [LMFBR]

Abstract

A nodal method is developed for the solution of the neutron-diffusion equation in two- and three-dimensional hexagonal geometries. The nodal scheme has been incorporated as an option in the finite-difference diffusion-theory code DIF3D, and is intended for use in the analysis of current LMFBR designs. The nodal equations are derived using higher-order polynomial approximations to the spatial dependence of the flux within the hexagonal-z node. The final equations, which are cast in the form of inhomogeneous response-matrix equations for each energy group, involved spatial moments of the node-interior flux distribution plus surface-averaged partial currents across the faces of the node. These equations are solved using a conventional fission-source iteration accelerated by coarse-mesh rebalance and asymptotic source extrapolation. This report describes the mathematical development and numerical solution of the nodal equations, as well as the use of the nodal option and details concerning its programming structure. This latter information is intended to supplement the information provided in the separate documentation of the DIF3D code.

Authors:
Publication Date:
Research Org.:
Argonne National Lab., IL (USA)
OSTI Identifier:
6318594
Report Number(s):
ANL-83-1
ON: DE83011019
DOE Contract Number:
W-31-109-ENG-38
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; COMPUTER CODES; D CODES; LMFBR TYPE REACTORS; REACTOR CORES; REACTOR KINETICS; NEUTRON FLUX; REACTIVITY COEFFICIENTS; FUEL ASSEMBLIES; NEUTRON DIFFUSION EQUATION; THREE-DIMENSIONAL CALCULATIONS; TWO-DIMENSIONAL CALCULATIONS; BREEDER REACTORS; DIFFERENTIAL EQUATIONS; EPITHERMAL REACTORS; EQUATIONS; FAST REACTORS; FBR TYPE REACTORS; KINETICS; LIQUID METAL COOLED REACTORS; RADIATION FLUX; REACTOR COMPONENTS; REACTORS; 210500* - Power Reactors, Breeding

Citation Formats

Lawrence, R.D. DIF3D nodal neutronics option for two- and three-dimensional diffusion theory calculations in hexagonal geometry. [LMFBR]. United States: N. p., 1983. Web. doi:10.2172/6318594.
Lawrence, R.D. DIF3D nodal neutronics option for two- and three-dimensional diffusion theory calculations in hexagonal geometry. [LMFBR]. United States. doi:10.2172/6318594.
Lawrence, R.D. Tue . "DIF3D nodal neutronics option for two- and three-dimensional diffusion theory calculations in hexagonal geometry. [LMFBR]". United States. doi:10.2172/6318594. https://www.osti.gov/servlets/purl/6318594.
@article{osti_6318594,
title = {DIF3D nodal neutronics option for two- and three-dimensional diffusion theory calculations in hexagonal geometry. [LMFBR]},
author = {Lawrence, R.D.},
abstractNote = {A nodal method is developed for the solution of the neutron-diffusion equation in two- and three-dimensional hexagonal geometries. The nodal scheme has been incorporated as an option in the finite-difference diffusion-theory code DIF3D, and is intended for use in the analysis of current LMFBR designs. The nodal equations are derived using higher-order polynomial approximations to the spatial dependence of the flux within the hexagonal-z node. The final equations, which are cast in the form of inhomogeneous response-matrix equations for each energy group, involved spatial moments of the node-interior flux distribution plus surface-averaged partial currents across the faces of the node. These equations are solved using a conventional fission-source iteration accelerated by coarse-mesh rebalance and asymptotic source extrapolation. This report describes the mathematical development and numerical solution of the nodal equations, as well as the use of the nodal option and details concerning its programming structure. This latter information is intended to supplement the information provided in the separate documentation of the DIF3D code.},
doi = {10.2172/6318594},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Tue Mar 01 00:00:00 EST 1983},
month = {Tue Mar 01 00:00:00 EST 1983}
}

Technical Report:

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  • The mathematical development and numerical solution of the finite-difference equations are summarized. The report provides a guide for user application and details the programming structure of DIF3D. Guidelines are included for implementing the DIF3D export package on several large scale computers. Optimized iteration methods for the solution of large-scale fast-reactor finite-difference diffusion theory calculations are presented, along with their theoretical basis. The computational and data management considerations that went into their formulation are discussed. The methods utilized include a variant of the Chebyshev acceleration technique applied to the outer fission source iterations and an optimized block successive overrelaxation method formore » the within-group iterations. A nodal solution option intended for analysis of LMFBR designs in two- and three-dimensional hexagonal geometries is incorporated in the DIF3D package and is documented in a companion report, ANL-83-1.« less
  • A nodal method is developed for the solution of the multigroup neutron-diffusion equation in three-dimensional hexagonal-z geometry. The method employs an extension to hexagonal geometry of the transverse-integration procedure used extensively in the development of nodal schemes in Cartesian geometry. The partially-integrated fluxes in the three hex-plane directions are approximated by a polynomial tailored to the unique properties of the transverse-integrated equations in hexagonal geometry. The final equations, which are cast in the form of local inhomogeneous response matrix equations for each energy group, involve spatial moments of the node-interior flux distribution plus surface-averaged partial currents across the faces ofmore » the node.« less
  • Advanced nodal methods for the solution of the multigroup neutron diffusion and transport theory equations in three-dimensional hexagonal-z geometry are described. The code HEXNOD allows an accurate and efficient calculation of three-dimensional problems for fast reactors and high converter light water reactors. A unique capability of HEXNOD is the accurate solution of global three-dimensional neutron transport problems for fast reactors with very small computing times. The accuracy of the nodal diffusion and transport approximations is demonstrated by comparison with conventional finite difference methods and Monte Carlo calculations for a number of mathematical benchmark problems. Based on numerical results, it ismore » concluded that the code HEXNOD is well suited for three-dimensional routine analysis of fast reactors and, in particular, as the neutronics module of the generalized quasi-static kinetics program HEXNODYN, which is currently being developed as part of the European accident code EAC-2.« less
  • This report documents the computer code VALE designed to solve multigroup neutronics problems with the diffusion theory approximation to neutron transport for a triagonal arrangement of mesh points on planes in two- and three-dimensional geometry. This code parallels the VENTURE neutronics code in the local computation system, making exposure and fuel management capabilities available. It uses and generates interface data files adopted in the cooperative effort sponsored by Reactor Physics RRT Division of the US DOE. The programming in FORTRAN is straightforward, although data is transferred in blocks between auxiliary storage devices and main core, and direct access schemes aremore » used. The size of problems which can be handled is essentially limited only by cost of calculation since the arrays are variably dimensioned. The memory requirement is held down while data transfer during iteration is increased only as necessary with problem size. There is provision for the more common boundary conditions including the repeating boundary, 180/sup 0/ rotational symmetry, and the rotational symmetry conditions for the 30/sup 0/, 60/sup 0/, and 120/sup 0/ triangular grids on planes. A variety of types of problems may be solved: the usual neutron flux eignevalue problem, or a direct criticality search on the buckling, on a reciprocal velocity absorber (prompt mode), or on nuclide concentrations. The adjoint problem and fixed source problem may be solved, as well as the dominating higher harmonic, or the importance problem for an arbitrary fixed source.« less
  • DIF3D solves multigroup diffusion theory eigenvalue, adjoint, fixed source, and criticality (concentration, buckling, and dimension search) problems in 1, 2, and 3-space dimensions for orthogonal (rectangular or cylindrical), triangular, and hexagonal geometries. Anisotropic diffusion theory coefficients are permitted. Flux and power density maps by mesh cell and regionwise balance integrals are provided. Although primarily designed for fast reactor problems, upscattering and internal black boundary conditions are also treated.