Abstract
| Original language | English |
|---|---|
| Pages (from-to) | 104-116 |
| Journal | Annals of Nuclear Energy |
| Volume | 78 |
| DOIs | |
| Publication status | Published - 2015 |
| MoE publication type | A1 Journal article-refereed |
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Keywords
- Serpent
- MOnte Carlo
- directional diffusion coefficient
- TRIZ
- SFR
- critically spectrum
Cite this
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Monte Carlo current-based diffusion coefficients : Application to few-group constants generation in Serpent. / Dorval, E.; Leppänen, J.
In: Annals of Nuclear Energy, Vol. 78, 2015, p. 104-116.Research output: Contribution to journal › Article › Scientific › peer-review
TY - JOUR
T1 - Monte Carlo current-based diffusion coefficients
T2 - Application to few-group constants generation in Serpent
AU - Dorval, E.
AU - Leppänen, J.
PY - 2015
Y1 - 2015
N2 - 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.
AB - 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.
KW - Serpent
KW - MOnte Carlo
KW - directional diffusion coefficient
KW - TRIZ
KW - SFR
KW - critically spectrum
U2 - 10.1016/j.anucene.2014.12.011
DO - 10.1016/j.anucene.2014.12.011
M3 - Article
VL - 78
SP - 104
EP - 116
JO - Annals of Nuclear Energy
JF - Annals of Nuclear Energy
SN - 0306-4549
ER -