3-D extension C5G7 MOX benchmark calculation using threedant code
- 1. Hanyang Univ., Dept. of Nuclear Engineering, Seoul (Korea, Republic of)
- 2. OECD NEA, Le Seine Saint-Germain, 92 - Issy les Moulineaux (France)
Description
It pursued the benchmark on deterministic 3-D MOX fuel assembly transport calculations without spatial homogenization (C5G7 MOX Benchmark Extension). The goal of this benchmark is to provide a more through test results for the abilities of current available 3-D methods to handle the spatial heterogeneities of reactor core. The benchmark requires solutions in the form of normalized pin powers as well as the eigenvalue for each of the control rod configurations; without rod, with A rods, and with B rods. In this work, the DANTSYS code package was applied to analyze the 3-D Extension C5G7 MOX Benchmark problems. The THREEDANT code within the DANTSYS code package, which solves the 3-D transport equation in x-y-z, and r-z-theta geometries, was employed to perform the benchmark calculations. To analyze the benchmark with the THREEDANT code, proper spatial and angular approximations were made. Several calculations were performed to investigate the effects of the different spatial approximations on the accuracy. The results from these sensitivity studies were analyzed and discussed. From the results, it is found that the 4*4 grid per pin cell is sufficiently refined so that very little benefit is obtained by increasing the mesh size. (authors)
Availability note (English)
Available from SFEN, 5 rue des Morillons, 75015 - Paris (France)Additional details
Publishing Information
- Publisher
- SFEN
- Imprint Place
- Paris (France)
- Imprint Pagination
- 12 p.
- Report number
- INIS-FR--09-1013
Conference
- Title
- international topical meeting on mathematics and computation, supercomputing, reactor physics and nuclear and biological applications
- Acronym
- M and C 2005
- Dates
- 12-15 Sep 2005
- Place
- Avignon (France)
INIS
- Country of Publication
- France
- Country of Input or Organization
- France
- INIS RN
- 40084869
- Subject category
- S22: GENERAL STUDIES OF NUCLEAR REACTORS;
- Resource subtype / Literary indicator
- Conference, Non-conventional Literature
- Descriptors DEI
- BENCHMARKS; D CODES; FUEL ASSEMBLIES; MESH GENERATION; MIXED OXIDE FUELS; NEUTRON TRANSPORT THEORY; T CODES; THREE-DIMENSIONAL CALCULATIONS; VALIDATION
- Descriptors DEC
- COMPUTER CODES; ENERGY SOURCES; FUELS; MATERIALS; NUCLEAR FUELS; REACTOR MATERIALS; SOLID FUELS; TESTING; TRANSPORT THEORY
Optional Information
- Notes
- 7 refs.