Monte Carlo current-based diffusion coefficients: Application to few-group constants generation in Serpent

E. Dorval, J. Leppänen

Research output: Contribution to journalArticleScientificpeer-review

8 Citations (Scopus)

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 languageEnglish
Pages (from-to)104-116
JournalAnnals of Nuclear Energy
Volume78
DOIs
Publication statusPublished - 2015
MoE publication typeA1 Journal article-refereed

Fingerprint

Sodium
Reactor cores
Fast reactors
Neutrons
Monte Carlo methods
Boundary conditions
Vacuum

Keywords

  • Serpent
  • MOnte Carlo
  • directional diffusion coefficient
  • TRIZ
  • SFR
  • critically spectrum

Cite this

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title = "Monte Carlo current-based diffusion coefficients: Application to few-group constants generation in Serpent",
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.",
keywords = "Serpent, MOnte Carlo, directional diffusion coefficient, TRIZ, SFR, critically spectrum",
author = "E. Dorval and J. Lepp{\"a}nen",
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language = "English",
volume = "78",
pages = "104--116",
journal = "Annals of Nuclear Energy",
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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 journalArticleScientificpeer-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 -