Monte Carlo methodologies for neutron streaming in diffusion calculations

Application to directional diffusion coefficients and leakage models in XS generation: Dissertation

Research output: ThesisDissertationCollection of Articles

Abstract

Neutron transport calculations by Monte Carlo methods are finding increased application in nuclear reactor simulations. In particular, a versatile approach entails the use of a 2-step pro-cedure, with Monte Carlo as a few-group cross section data generator at lattice level, followed by deterministic multi-group diffusion calculations at core level. In this thesis, the Serpent 2 Monte Carlo reactor physics burnup calculation code is used in order to test a set of diffusion coefficient models, as well as neutron leakage methodologies at assembly level. The tests include novel anisotropic diffusion coefficient and heterogeneous leakage models developed and implemented by the author. The analyses are mainly focused on a sodium-cooled fast reactor system, for which few-group cross section data was generated by stochastic methods with Serpent 2. The quality of the full-core diffusion results is evaluated by contrasting system eigenvalues and power distributions against detailed, full-core reference solutions also supplied by the Serpent 2 code and the same nuclear data library. Whereas the new anisotropic diffusion coefficient formalism exhibits improved performance in the fast reactor system studied, there are restrictions to its applicability in other reactor de-signs. The newly proposed leakage model has a similar performance to that one of albedo ite-rations, and provides valuable information about the space-energy coupling of the scalar neu-tron flux at lattice level. This hitherto unavailable information does not entail a significant computational cost. In sodium-cooled fast reactor calculations, the quality of diffusion theory results can be im-proved by either using directional diffusion coefficients and a fine energy mesh, or via leakage-corrected discontinuity factors. These factors can be calculated using net neutron currents supplied by heterogeneous leakage models. Preliminary results from this research also suggest that the studies maybe extended to graphite-moderated, gas-cooled reactors.
Original languageEnglish
QualificationDoctor Degree
Awarding Institution
  • Aalto University
Supervisors/Advisors
  • Tuomisto, Filip, Supervisor, External person
  • Leppänen, Jaakko, Advisor
Award date18 May 2016
Place of PublicationEspoo
Publisher
Print ISBNs978-952-60-6735-3
Electronic ISBNs978-952-60-6736-0
Publication statusPublished - 2016
MoE publication typeG5 Doctoral dissertation (article)

Fingerprint

leakage
diffusion coefficient
methodology
neutrons
reactors
reactor physics
sodium
gas cooled reactors
rations
diffusion theory
theses
nuclear reactors
cross sections
albedo
Monte Carlo method
mesh
constrictions
discontinuity
eigenvalues
generators

Keywords

  • Monte Carlo
  • Serpent
  • directional diffusion coefficient
  • TRIZ
  • TRIVAC
  • neutron leakage
  • layer expansion
  • albedo
  • B1
  • discontinuity factor
  • SFR

Cite this

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title = "Monte Carlo methodologies for neutron streaming in diffusion calculations: Application to directional diffusion coefficients and leakage models in XS generation: Dissertation",
abstract = "Neutron transport calculations by Monte Carlo methods are finding increased application in nuclear reactor simulations. In particular, a versatile approach entails the use of a 2-step pro-cedure, with Monte Carlo as a few-group cross section data generator at lattice level, followed by deterministic multi-group diffusion calculations at core level. In this thesis, the Serpent 2 Monte Carlo reactor physics burnup calculation code is used in order to test a set of diffusion coefficient models, as well as neutron leakage methodologies at assembly level. The tests include novel anisotropic diffusion coefficient and heterogeneous leakage models developed and implemented by the author. The analyses are mainly focused on a sodium-cooled fast reactor system, for which few-group cross section data was generated by stochastic methods with Serpent 2. The quality of the full-core diffusion results is evaluated by contrasting system eigenvalues and power distributions against detailed, full-core reference solutions also supplied by the Serpent 2 code and the same nuclear data library. Whereas the new anisotropic diffusion coefficient formalism exhibits improved performance in the fast reactor system studied, there are restrictions to its applicability in other reactor de-signs. The newly proposed leakage model has a similar performance to that one of albedo ite-rations, and provides valuable information about the space-energy coupling of the scalar neu-tron flux at lattice level. This hitherto unavailable information does not entail a significant computational cost. In sodium-cooled fast reactor calculations, the quality of diffusion theory results can be im-proved by either using directional diffusion coefficients and a fine energy mesh, or via leakage-corrected discontinuity factors. These factors can be calculated using net neutron currents supplied by heterogeneous leakage models. Preliminary results from this research also suggest that the studies maybe extended to graphite-moderated, gas-cooled reactors.",
keywords = "Monte Carlo, Serpent, directional diffusion coefficient, TRIZ, TRIVAC, neutron leakage, layer expansion, albedo, B1, discontinuity factor, SFR",
author = "Eric Dorval",
note = "BA2123 89 p. + app. 66 p.",
year = "2016",
language = "English",
isbn = "978-952-60-6735-3",
series = "Aalto University publication series: Doctoral Dissertations",
publisher = "Aalto University",
number = "60",
address = "Finland",
school = "Aalto University",

}

TY - THES

T1 - Monte Carlo methodologies for neutron streaming in diffusion calculations

T2 - Application to directional diffusion coefficients and leakage models in XS generation: Dissertation

AU - Dorval, Eric

N1 - BA2123 89 p. + app. 66 p.

PY - 2016

Y1 - 2016

N2 - Neutron transport calculations by Monte Carlo methods are finding increased application in nuclear reactor simulations. In particular, a versatile approach entails the use of a 2-step pro-cedure, with Monte Carlo as a few-group cross section data generator at lattice level, followed by deterministic multi-group diffusion calculations at core level. In this thesis, the Serpent 2 Monte Carlo reactor physics burnup calculation code is used in order to test a set of diffusion coefficient models, as well as neutron leakage methodologies at assembly level. The tests include novel anisotropic diffusion coefficient and heterogeneous leakage models developed and implemented by the author. The analyses are mainly focused on a sodium-cooled fast reactor system, for which few-group cross section data was generated by stochastic methods with Serpent 2. The quality of the full-core diffusion results is evaluated by contrasting system eigenvalues and power distributions against detailed, full-core reference solutions also supplied by the Serpent 2 code and the same nuclear data library. Whereas the new anisotropic diffusion coefficient formalism exhibits improved performance in the fast reactor system studied, there are restrictions to its applicability in other reactor de-signs. The newly proposed leakage model has a similar performance to that one of albedo ite-rations, and provides valuable information about the space-energy coupling of the scalar neu-tron flux at lattice level. This hitherto unavailable information does not entail a significant computational cost. In sodium-cooled fast reactor calculations, the quality of diffusion theory results can be im-proved by either using directional diffusion coefficients and a fine energy mesh, or via leakage-corrected discontinuity factors. These factors can be calculated using net neutron currents supplied by heterogeneous leakage models. Preliminary results from this research also suggest that the studies maybe extended to graphite-moderated, gas-cooled reactors.

AB - Neutron transport calculations by Monte Carlo methods are finding increased application in nuclear reactor simulations. In particular, a versatile approach entails the use of a 2-step pro-cedure, with Monte Carlo as a few-group cross section data generator at lattice level, followed by deterministic multi-group diffusion calculations at core level. In this thesis, the Serpent 2 Monte Carlo reactor physics burnup calculation code is used in order to test a set of diffusion coefficient models, as well as neutron leakage methodologies at assembly level. The tests include novel anisotropic diffusion coefficient and heterogeneous leakage models developed and implemented by the author. The analyses are mainly focused on a sodium-cooled fast reactor system, for which few-group cross section data was generated by stochastic methods with Serpent 2. The quality of the full-core diffusion results is evaluated by contrasting system eigenvalues and power distributions against detailed, full-core reference solutions also supplied by the Serpent 2 code and the same nuclear data library. Whereas the new anisotropic diffusion coefficient formalism exhibits improved performance in the fast reactor system studied, there are restrictions to its applicability in other reactor de-signs. The newly proposed leakage model has a similar performance to that one of albedo ite-rations, and provides valuable information about the space-energy coupling of the scalar neu-tron flux at lattice level. This hitherto unavailable information does not entail a significant computational cost. In sodium-cooled fast reactor calculations, the quality of diffusion theory results can be im-proved by either using directional diffusion coefficients and a fine energy mesh, or via leakage-corrected discontinuity factors. These factors can be calculated using net neutron currents supplied by heterogeneous leakage models. Preliminary results from this research also suggest that the studies maybe extended to graphite-moderated, gas-cooled reactors.

KW - Monte Carlo

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KW - directional diffusion coefficient

KW - TRIZ

KW - TRIVAC

KW - neutron leakage

KW - layer expansion

KW - albedo

KW - B1

KW - discontinuity factor

KW - SFR

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ER -