COMET solutions to whole core CANDU-6 benchmark problems
- Nuclear and Radiological Engineering / Medical Physics Programs, George W. Woodruff School, Georgia Inst. of Technology, Atlanta, GA 30332-0405 (United States)
In this paper, the coarse mesh transport code COMET is used to solve CANDU-6 benchmark problems in two and three dimensional geometry. These problems are representative of a simplified quarter core reactor model. The COMET solutions, the core eigenvalue and the fuel pin fission density distribution, are compared to those from the Monte Carlo code MCNP using two-group cross sections. COMET decomposes the core volume into a set of non-overlapping sub-volumes (coarse meshes) and uses pre-computed heterogeneous response functions that are constructed using Legendre polynomials as boundary conditions to generate a user selected whole core solution (e.g., the core eigenvalue and fuel pin fission density distribution). These response functions are pre-computed by performing fixed source calculations with a modified version of MCNP in only the unique coarse meshes in the core. Reference solutions are calculated by MCNP5 with a two-group energy library generated with the HELIOS lattice code. In the 2-D problem, the angular current on the coarse mesh interfaces in COMET is expanded to 2. order in both spatial and angular variables. The COMET eigenvalue error is 0.09%. The corresponding average error in the fission density over all 3515 fuel pins is 0.5%. The maximum error observed is 2.0%. For the 3-D case, with 4. order expansion in space and azimuthal angle and 2. order expansion in the cosine of the polar angle, the eigenvalue differs from the reference solution by 0.05%. The average fission density error over the 42180 fuel pins is 0.7% with a maximum error of 3.3%. (authors)
- Research Organization:
- American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
- OSTI ID:
- 22039519
- Resource Relation:
- Conference: PHYSOR-2006: American Nuclear Society's Topical Meeting on Reactor Physics - Advances in Nuclear Analysis and Simulation, Vancouver, BC (Canada), 10-14 Sep 2006; Other Information: Country of input: France; 11 refs.
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BENCHMARKS
BOUNDARY CONDITIONS
CANDU TYPE REACTORS
COMPUTERIZED SIMULATION
CROSS SECTIONS
DENSITY
EIGENFUNCTIONS
EIGENVALUES
ERRORS
FISSION
FUEL PINS
LEGENDRE POLYNOMIALS
MATHEMATICAL SOLUTIONS
MONTE CARLO METHOD
REACTOR CORES
RESPONSE FUNCTIONS
THREE-DIMENSIONAL CALCULATIONS
TRANSPORT THEORY