Collision probability methods - An update

Conference · Sun Jan 01 00:00:00 EST 1989 · Transactions of the American Nuclear Society; (USA)

OSTI ID:6529305

Sanchez, R ^[1]

---

1. CEA/CEN, Saclay (France)

A preliminary step in reactor coarse-mesh calculations is the preparation of an adequate macroscopic cross-section library by spatial smearing and group collapsing of detailed cell and/or assembly multigroup transport calculations. Currently, only two different techniques are regularly used for the determination of the multigroup transport flux in heterogeneous cells or assemblies: (1) methods based on the discretization of the integrodifferential transport equation and (2) collision probability methods for the numerical solution of the integral transport equation. Collision probability (CP) methods are derived either from the direct use of a quadrature formula or by the expansion of the unknown scalar flux in terms of representation functions. The new computer generation, with faster processing and a larger central memory that can support precise tabulations of transcendental functions, has considerably reduced the cost of multidimensional CP calculations. CP methods have come of age and are currently used in bidimensional routine reactor calculations through the world. Recent developments in CP methods include double-heterogeneity approximations and multilevel methods. Current CP research in France is oriented toward enhanced vectorization and to the use of object-oriented languages for the development of general-purpose multidimensional tracking subroutines.

• OSTI ID: 6529305
• Report Number(s): CONF-891103-; CODEN: TANSA; TRN: 90-035187
• Journal Information: Transactions of the American Nuclear Society; (USA), Vol. 60; Conference: Winter meeting of the American Nuclear Society (ANS) and nuclear power and technology exhibit, San Francisco, CA (USA), 26-30 Nov 1989; ISSN 0003-018X
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
REACTOR KINETICS
COLLISION INTEGRALS
ANISOTROPY
CROSS SECTIONS
FINITE ELEMENT METHOD
GROUP THEORY
MANY-DIMENSIONAL CALCULATIONS
MATHEMATICS
NEUTRON FLUX
PROBABILITY
PWR TYPE REACTORS
SCALARS
SCATTERING
SPACE DEPENDENCE
VECTORS
INTEGRALS
KINETICS
NUMERICAL SOLUTION
RADIATION FLUX
REACTORS
TENSORS
WATER COOLED REACTORS
WATER MODERATED REACTORS
220100* - Nuclear Reactor Technology- Theory & Calculation