Comet solutions to a stylized BWR benchmark problem

Conference · Sun Jul 01 00:00:00 EDT 2012

OSTI ID:22105854

Zhang, D.; Rahnema, F. ^[1]

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1. Georgia Inst. of Technology, 770 State Street, Atlanta, GA 30332-0745 (United States)

In this paper, a stylized 3-D BWR benchmark problem was used to evaluate the performance of the coarse mesh radiation transport method COMET. The benchmark problem consists of 560 fuel bundles at 3 different burnups and 3 coolant void states. The COMET solution was compared with the corresponding Monte Carlo reference solution using the same 2-group material cross section library for three control blade (rod) configurations, namely, all rods out (ARO), all rods in (ARI) and some rods in (SRJ). The differences in the COMET and MCNP eigenvalues were 43 pcm, 66 pcm and 32 pcm for the ARO, ARI and SRI cases, respectively. These differences are all within 3 standard deviations of the COMET uncertainty. The average relative differences in the bundle averaged fission densities for these three cases were 0.89%, 1.24%, and 1.05%, respectively. The corresponding differences in the fuel pin averaged fission densities were 1.24%, 1.84% and 1.29%, respectively. It was found that COMET is 3,000 times faster than Monte Carlo, while its statistical uncertainty in the fuel pin fission density is much lower than that of Monte Carlo (i.e., {approx}40 times lower). (authors)

• Research Organization: American Nuclear Society, Inc., 555 N. Kensington Avenue, La Grange Park, Illinois 60526 (United States)
• OSTI ID: 22105854
• Resource Relation: Conference: PHYSOR 2012: Conference on Advances in Reactor Physics - Linking Research, Industry, and Education, Knoxville, TN (United States), 15-20 Apr 2012; Other Information: Country of input: France; 7 refs.
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
BENCHMARKS
BURNUP
BWR TYPE REACTORS
CONTROL ELEMENTS
COOLANTS
EIGENFUNCTIONS
EIGENVALUES
FISSION
FUEL ELEMENT CLUSTERS
FUEL PINS
MONTE CARLO METHOD
NUCLEAR DATA COLLECTIONS
RADIATION TRANSPORT
REACTOR CORES
SIMULATION
THREE-DIMENSIONAL CALCULATIONS
TRANSPORT THEORY
VOIDS