Published April 1993
| Version v1
Miscellaneous
Benchmark calculations of power distribution within assemblies
- 1. EDF, Direction des Etudes et Recherches, 92 - Clamart (France)
- 2. EDF, Direction de l'Equipement, 92 - Clamart (France)
- 3. EDF, Direction Production Transport, 92 - Clamart (France)
Description
This paper presents the analysis and intercomparison of results produced in the framework of the international OECD Benchmark L-336, related to pin power distribution assessment within 'heterogeneous core type' configurations. The major objective of the benchmark can be considered as achieved. A fairly good convergence of the results is obtained although different techniques were used: Diffusion Theory with different numerical schemes: Finite Differences, Nodal, or Transport Theory with Pij, Sn, or Monte Carlo schemes applied to different kinds of nodalization (one or several meshes per element, an element being or a pin for the pin by pin calculations, either an assembly for the homogeneous calculations). (orig./DG)
Additional details
Publishing Information
- ISBN
- 3-923704-11-9
- Imprint Title
- Mathematical methods and supercomputing in nuclear applications. Proceedings. Vol. 2
- Imprint Pagination
- 822 p.
- Journal Page Range
- p. 259-271.
Conference
- Title
- Joint international conference on mathematical methods and supercomputing in nuclear applications (M and C and SNA '93).
- Dates
- 19-23 Apr 1993.
- Place
- Karlsruhe (Germany).
INIS
- Country of Publication
- Germany
- Country of Input or Organization
- Germany
- INIS RN
- 25062070
- Subject category
- S22: GENERAL STUDIES OF NUCLEAR REACTORS;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- BENCHMARKS; FINITE DIFFERENCE METHOD; FUEL RODS; HETEROGENEOUS REACTOR CORES; HOMOGENEOUS REACTORS; INTERLABORATORY COMPARISONS; MESH GENERATION; MIXED OXIDE FUELS; MONTE CARLO METHOD; NEUTRON DIFFUSION EQUATION; NEUTRON TRANSPORT THEORY; NODAL EXPANSION METHOD; POWER DISTRIBUTION; URANIUM DIOXIDE
- Descriptors DEC
- ACTINIDE COMPOUNDS; CALCULATION METHODS; CHALCOGENIDES; DISTRIBUTION; ENERGY SOURCES; FUEL ELEMENTS; FUELS; ITERATIVE METHODS; MATERIALS; NUCLEAR FUELS; NUMERICAL SOLUTION; OXIDES; OXYGEN COMPOUNDS; REACTOR COMPONENTS; REACTOR CORES; REACTOR MATERIALS; REACTORS; SOLID FUELS; SPATIAL DISTRIBUTION; TRANSPORT THEORY; URANIUM COMPOUNDS; URANIUM OXIDES