Analysis of EBR-II Neutron and Photon Physics by Multidimensional Transport-Theory Techniques

Conference · Mon Apr 11 00:00:00 EDT 1994

OSTI ID:10132792

Jacqmin, R. P. ^[1]; Palmiotti, G. ^[2]; Finck, P. J. ^[1]

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1. Argonne National Lab. (ANL), Argonne, IL (United States)
2. CEA Centre d`Etudes de Cadarache, Saint-Paul-lez-Durance (France); Argonne National Lab. (ANL), Argonne, IL (United States)

This paper contains a review of the challenges specific to the EBR-II core physics, a description of the methods and techniques which have been developed for addressing these challenges, and the results of some validation studies relative to power-distribution calculations. Numerical tests have shown that the VARIANT nodal code yields eigenvalue and power predictions as accurate as finite difference and discrete ordinates transport codes, at a small fraction of the cost. Comparisons with continuous-energy Monte Carlo results have proven that the errors introduced by the use of the diffusion-theory approximation in the collapsing procedure to obtain broad-group cross sections, kerma factors, and photon-production matrices, have a small impact on the EBR-II neutron/photon power distribution.

• Research Organization: Argonne National Lab. (ANL), Argonne, IL (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP); USDOE Nuclear Energy Programs
• DOE Contract Number: W-31109-ENG-38
• OSTI ID: 10132792
• Report Number(s): ANL/RA/CP-80649; CONF-940407-15; ON: DE94007640; TRN: 94:007388
• Resource Relation: Conference: American Nuclear Society Topical Meeting on Advances in Reactor Physics, Knoxville, TN (United States), 11-14 Apr 1994; Other Information: PBD: [1994]
• Country of Publication: United States
• Language: English

Related Subjects

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS
97 MATHEMATICS AND COMPUTING
EBR-2 REACTOR
REACTOR PHYSICS
NEUTRONS
KERMA
PHOTON TRANSPORT
NEUTRON TRANSPORT
CRITICALITY
V CODES
POWER DISTRIBUTION
D CODES
T CODES
NEUTRON FLUX
CROSS SECTIONS
M CODES
NUCLEAR DATA COLLECTIONS
Nuclear Criticality Safety Program (NCSP)
Variational Nodal Method
Experimental Breeder Reactor-II (EBR-II)
Monte Carlo
TWODANT
Kerma Factors
210500
220100
POWER REACTORS
BREEDING
THEORY AND CALCULATION