Lattice physics capabilities of the SCALE code system using TRITON
Creators
- 1. Oak Ridge National Laboratory, MS 6170, P.O. Box 2008, Oak Ridge, TN 37831-6170 (United States)
Description
This paper describes ongoing calculations used to validate the TRITON depletion module in SCALE for light water reactor (LWR) fuel lattices. TRITON has been developed to provide improved resolution for lattice physics mixed-oxide fuel assemblies as programs to burn such fuel in the United States begin to come online. Results are provided for coupled TRITON/PARCS analyses of an LWR core in which TRITON was employed for generation of appropriately weighted few-group nodal cross-sectional sets for use in core-level calculations using PARCS. Additional results are provided for code-to-code comparisons for TRITON and a suite of other depletion packages in the modeling of a conceptual next-generation boiling water reactor fuel assembly design. Results indicate that the set of SCALE functional modules used within TRITON provide an accurate means for lattice physics calculations. Because the transport solution within TRITON provides a generalized-geometry capability, this capability is extensible to a wide variety of non-traditional and advanced fuel assembly designs. (authors)
Additional details
Publishing Information
- Publisher
- American Nuclear Society - ANS
- Imprint Place
- La Grange Park (United States)
- ISBN
- 0-89448-697-7
- Imprint Pagination
- 10 p.
Conference
- Title
- American Nuclear Society's Topical Meeting on Reactor Physics - Advances in Nuclear Analysis and Simulation
- Acronym
- PHYSOR-2006
- Dates
- 10-14 Sep 2006
- Place
- Vancouver, BC (Canada)
INIS
- Country of Publication
- United States
- Country of Input or Organization
- France
- INIS RN
- 43129884
- Subject category
- S21: SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; S97: MATHEMATICAL METHODS AND COMPUTING;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- BURNUP; BWR TYPE REACTORS; COMPARATIVE EVALUATIONS; DESIGN; FUEL ASSEMBLIES; GEOMETRY; MATHEMATICAL SOLUTIONS; MIXED OXIDE FUELS; NEUTRON TRANSPORT; REACTOR LATTICE PARAMETERS; RESOLUTION; S CODES; T CODES; THREE-DIMENSIONAL CALCULATIONS
- Descriptors DEC
- COMPUTER CODES; ENERGY SOURCES; ENRICHED URANIUM REACTORS; EVALUATION; FUELS; MATERIALS; MATHEMATICS; NEUTRAL-PARTICLE TRANSPORT; NUCLEAR FUELS; POWER REACTORS; RADIATION TRANSPORT; REACTOR MATERIALS; REACTORS; SOLID FUELS; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- 11 refs.