Abstract
Criticality eigenvalue and power distributions of a
medium-sized sodium-cooled fast reactor core were studied
by combined diffusion theory and Monte Carlo methods.
Few-group XS data sets generated by the Monte Carlo code
Serpent for both normal and sodium-voided cells were used
in the multi-group diffusion code TRIZ. Two different
approaches were adopted in order to account for neutron
leakage at fuel assembly level: a radially-reflected and
axially-heterogeneous model with vacuum boundary
conditions at the bottom and the top; and a more typical
infinite cell calculation, followed by criticality
spectrum corrections. In addition to the standard
diffusion coefficients calculated by Serpent, a novel
method for the calculation of directional diffusion
coefficients was implemented and tested, yielding
satisfactory results for normal and sodium-voided
conditions, using Monte Carlo results as a reference. The
feasibility of using the B1B1 criticality spectrum as a
weighting function for these diffusion coefficients was
also tested, and slightly better estimates of k-eff
were obtained when compared to the direct use of
diffusion coefficients arising from the solution of the
B1B1 equations.
Original language | English |
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Pages (from-to) | 104-116 |
Journal | Annals of Nuclear Energy |
Volume | 78 |
DOIs | |
Publication status | Published - 2015 |
MoE publication type | A1 Journal article-refereed |
Keywords
- Serpent
- MOnte Carlo
- directional diffusion coefficient
- TRIZ
- SFR
- critically spectrum