Neutron Flux Interpolation with Finite Element Method in the Nuclear Fuel Cell Calculation using Collision Probability Method

Journal Article · Thu Dec 23 00:00:00 EST 2010 · AIP Conference Proceedings

DOI:https://doi.org/10.1063/1.3537911· OSTI ID:21511433

Shafii, M Ali ^[1]; Su'ud, Zaki; Waris, Abdul; Kurniasih, Neny ^[1]; Ariani, Menik ^[2]; Yulianti, Yanti ^[3]

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1. Departmen of Physics Bandung Institute of Technology, Jl. Ganesha 10, Bandung 40134 (Indonesia)
2. Physics Department, Sriwijaya University, Kampus Indralaya, Ogan Ilir, Sumatera Selatan (Indonesia)
3. Physics Department, Lampung University, Jl.Sumantri Brojonegoro no 1, Lampung (Indonesia)

Nuclear reactor design and analysis of next-generation reactors require a comprehensive computing which is better to be executed in a high performance computing. Flat flux (FF) approach is a common approach in solving an integral transport equation with collision probability (CP) method. In fact, the neutron flux distribution is not flat, even though the neutron cross section is assumed to be equal in all regions and the neutron source is uniform throughout the nuclear fuel cell. In non-flat flux (NFF) approach, the distribution of neutrons in each region will be different depending on the desired interpolation model selection. In this study, the linear interpolation using Finite Element Method (FEM) has been carried out to be treated the neutron distribution. The CP method is compatible to solve the neutron transport equation for cylindrical geometry, because the angle integration can be done analytically. Distribution of neutrons in each region of can be explained by the NFF approach with FEM and the calculation results are in a good agreement with the result from the SRAC code. In this study, the effects of the mesh on the k{sub eff} and other parameters are investigated.

• OSTI ID: 21511433
• Journal Information: AIP Conference Proceedings, Vol. 1325, Issue 1; Conference: 4. Asian physics international symposium, Bandung, West Java (Indonesia), 12-13 Oct 2010; Other Information: DOI: 10.1063/1.3537911; (c) 2010 American Institute of Physics; ISSN 0094-243X
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
COLLISION PROBABILITY METHOD
COLLISIONS
COMPUTERIZED SIMULATION
CROSS SECTIONS
CYLINDRICAL CONFIGURATION
DESIGN
DISTRIBUTION
FINITE ELEMENT METHOD
FUEL CELLS
INTERPOLATION
NEUTRON FLUX
NEUTRON SOURCES
NEUTRON TRANSPORT THEORY
NEUTRONS
NUCLEAR FUELS
PROBABILITY
REACTORS
THERMONUCLEAR REACTORS
BARYONS
CALCULATION METHODS
CONFIGURATION
DIRECT ENERGY CONVERTERS
ELECTROCHEMICAL CELLS
ELEMENTARY PARTICLES
ENERGY SOURCES
FERMIONS
FUELS
HADRONS
MATERIALS
MATHEMATICAL SOLUTIONS
NUCLEONS
NUMERICAL SOLUTION
PARTICLE SOURCES
RADIATION FLUX
RADIATION SOURCES
REACTOR MATERIALS
SIMULATION
TRANSPORT THEORY