Advanced nodal neutron diffusion method with space-dependent cross sections: ILLICO-VX

Conference · Thu Jan 01 00:00:00 EST 1987 · Trans. Am. Nucl. Soc.; (United States)

OSTI ID:6831090

Rajic, H L; Ougouag, A M

Advanced transverse integrated nodal methods for neutron diffusion developed since the 1970s require that node- or assembly-homogenized cross sections be known. The underlying structural heterogeneity can be accurately accounted for in homogenization procedures by the use of heterogeneity or discontinuity factors. Other (milder) types of heterogeneity, burnup-induced or due to thermal-hydraulic feedback, can be resolved by explicitly accounting for the spatial variations of material properties. This can be done during the nodal computations via nonlinear iterations. The new method has been implemented in the code ILLICO-VX (ILLICO variable cross-section method). Numerous numerical tests were performed. As expected, the convergence rate of ILLICO-VX is lower than that of ILLICO, requiring approx. 30% more outer iterations per k/sub eff/ computation. The methodology has also been implemented as the NOMAD-VX option of the NOMAD, multicycle, multigroup, two- and three-dimensional nodal diffusion depletion code. The burnup-induced heterogeneities (space dependence of cross sections) are calculated during the burnup steps.

• Research Organization: Univ. of Illinois, Urbana (USA)
• OSTI ID: 6831090
• Report Number(s): CONF-8711195-; TRN: 88-031987
• Journal Information: Trans. Am. Nucl. Soc.; (United States), Vol. 55; Conference: American Nuclear Society winter meeting, Los Angeles, CA, USA, 15 Nov 1987
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
REACTOR PHYSICS
I CODES
NEUTRON DIFFUSION EQUATION
ACCURACY
BENCHMARKS
BURNUP
CROSS SECTIONS
GROUP THEORY
HEAT TRANSFER
HYDRAULICS
MULTIPLICATION FACTORS
NONLINEAR PROBLEMS
SPACE DEPENDENCE
COMPUTER CODES
DIFFERENTIAL EQUATIONS
ENERGY TRANSFER
EQUATIONS
FLUID MECHANICS
MATHEMATICS
MECHANICS
PHYSICS
220100* - Nuclear Reactor Technology- Theory & Calculation