Recent developments in multidimensional transport. Methods for the APOLLO 2 lattice code
Description
A usual method for preparation of homogenized cross sections for reactor coarse mesh calculations is based on two dimensional multigroup transport treatment of an assembly together with an appropriate leakage model and reaction-rate-preserving homogenization technique. The actual generation of assembly spectrum codes based on collision probability methods is capable to treat complex geometries (i.e. irregular meshes of arbitrary shape), avoiding thus the modeling error which was introduced in codes with traditional tracking routines. The power and architecture of nowadays computers allow to treat spatial domains comprising several mutually-interacting assemblies using fine multigroup structure and keeping all geometrical details of interest. Increasing safety requirements demand detailed 2D and 3D calculations for very heterogeneous problems such as in control rod positioning, broken Pyrex rods, irregular compacting of MOX pellets at a MOX-UO2 interface and many others. An effort is been made to include accurate multidimensional transport methods in the APOLLO 2 lattice code. These include the extension to 3D axially-symmetric geometries of the general-geometry collision probability module TDT2 and the development of new 2D-3D characteristics methods for regular Cartesian meshes. Here we discuss the main features of recently developed multidimensional methods that are currently being tested. (authors). 10 refs., 1 figs., 1 tab
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Additional details
Publishing Information
- Imprint Pagination
- 6 p.
- Report number
- CEA-CONF--12086
Conference
- Title
- Annual meeting of the American Nuclear Society (ANS).
- Dates
- 25-29 Jun 1995.
- Place
- Philadelphia, PA (United States).
INIS
- Country of Publication
- France
- Country of Input or Organization
- France
- INIS RN
- 27047950
- Subject category
- S22: GENERAL STUDIES OF NUCLEAR REACTORS;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- A CODES; LEGENDRE POLYNOMIALS; MESH GENERATION; MIXED OXIDE FUELS; NEUTRON TRANSPORT; REACTOR CORES; THREE-DIMENSIONAL CALCULATIONS
- Descriptors DEC
- COMPUTER CODES; ENERGY SOURCES; FUELS; FUNCTIONS; MATERIALS; NEUTRAL-PARTICLE TRANSPORT; NUCLEAR FUELS; POLYNOMIALS; RADIATION TRANSPORT; REACTOR COMPONENTS; REACTOR MATERIALS; SOLID FUELS