Surface flux and partial current reconstruction method in integral transport theory

Conference · Fri Jan 01 00:00:00 EST 1993 · Transactions of the American Nuclear Society; (United States)

OSTI ID:5928210

Vujic, J L ^[1]

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1. Univ. of California, Berkeley (United States)

Regardless of recent advances in computer architectures, various nodal diffusion theory methods are still used for the global reactor-core calculations. It is known that the nodal diffusion theory solution usually cannot accurately predict the real neutronic behavior. Fortunately, various methods based on the equivalence theory have been developed in order to force the nodal diffusion schemes to reproduce the higher order transport theory results. A set of the discontinuity factors is defined for each node and calculated by some higher order transport theory method. Although the definitions of the discontinuity factors vary for different reactor types and different nodal schemes, the following quantities are usually required: node-averaged flux and sources, average net currents, partial currents, and average surface fluxes on each face of the node. In this paper, we present a newly developed method to determine the surface fluxes and partial currents on the internal and external domain boundaries, based on the exact collision/transfer probability (CTP) method.

• OSTI ID: 5928210
• Report Number(s): CONF-930601-; CODEN: TANSAO
• Journal Information: Transactions of the American Nuclear Society; (United States), Vol. 68; Conference: American Nuclear Society (ANS) annual meeting, San Diego, CA (United States), 20-24 Jun 1993; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

73 NUCLEAR PHYSICS AND RADIATION PHYSICS
22 GENERAL STUDIES OF NUCLEAR REACTORS
NEUTRON DIFFUSION EQUATION
NUMERICAL SOLUTION
NEUTRON TRANSPORT
COMPUTERIZED SIMULATION
NODAL EXPANSION METHOD
PARALLEL PROCESSING
REACTOR CORES
TRANSPORT THEORY
CALCULATION METHODS
DIFFERENTIAL EQUATIONS
EQUATIONS
NEUTRAL-PARTICLE TRANSPORT
PROGRAMMING
RADIATION TRANSPORT
REACTOR COMPONENTS
SIMULATION
663610* - Neutron Physics- (1992-)
220100 - Nuclear Reactor Technology- Theory & Calculation