The AN neutron transport by nodal diffusion

Conference · Mon Jul 01 00:00:00 EDT 2013

OSTI ID:22212695

Barbarino, A. ^[1]; Tomatis, D. ^[2]

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1. Politecnico di Torino, Corso Duca degli Abruzzi 24, 10129 Torino (Italy)
2. AREVA NP, Tour AREVA, 92084 Paris La Defense Cedex (France)

The two group diffusion model combined to a nodal approach in space is the preferred scheme for the industrial simulation of nuclear water reactors. The main selling point is the speed of computation, allowing a large number of parametric studies. Anyway, the drawbacks of the underlying diffusion equation may arise with highly heterogeneous interfaces, often encountered in modern UO{sub 2} and MO{sub x} fuel loading patterns, and boron less controlled systems. This paper aims at showing how the simplified AN transport model, equivalent to the well known SPN, can be implemented in standard diffusion codes with minor modifications. Some numerical results are illustrated. (authors)

• Research Organization: American Nuclear Society, 555 North Kensington Avenue, La Grange Park, IL 60526 (United States)
• OSTI ID: 22212695
• Resource Relation: Conference: M and C 2013: 2013 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, Sun Valley, ID (United States), 5-9 May 2013; Other Information: Country of input: France; 18 refs.; Related Information: In: Proceedings of the 2013 International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering - M and C 2013| 3016 p.
• Country of Publication: United States
• Language: English

Related Subjects

97 MATHEMATICAL METHODS AND COMPUTING
CALCULATION METHODS
COMPUTERIZED SIMULATION
DIFFUSION EQUATIONS
MIXED OXIDE FUELS
NEUTRON TRANSPORT
PARAMETRIC ANALYSIS
TRANSPORT THEORY
URANIUM DIOXIDE