Abstract
Original language  English 

Qualification  Doctor Degree 
Awarding Institution 

Supervisors/Advisors 

Award date  18 May 2016 
Place of Publication  Espoo 
Publisher  
Print ISBNs  9789526067353 
Electronic ISBNs  9789526067360 
Publication status  Published  2016 
MoE publication type  G5 Doctoral dissertation (article) 
Fingerprint
Keywords
 Monte Carlo
 Serpent
 directional diffusion coefficient
 TRIZ
 TRIVAC
 neutron leakage
 layer expansion
 albedo
 B1
 discontinuity factor
 SFR
Cite this
}
Monte Carlo methodologies for neutron streaming in diffusion calculations : Application to directional diffusion coefficients and leakage models in XS generation: Dissertation. / Dorval, Eric.
Espoo : Aalto University, 2016. 155 p.Research output: Thesis › Dissertation › Collection of Articles
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 2step procedure, with Monte Carlo as a fewgroup cross section data generator at lattice level, followed by deterministic multigroup 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 sodiumcooled fast reactor system, for which fewgroup cross section data was generated by stochastic methods with Serpent 2. The quality of the fullcore diffusion results is evaluated by contrasting system eigenvalues and power distributions against detailed, fullcore 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 designs. The newly proposed leakage model has a similar performance to that one of albedo iterations, and provides valuable information about the spaceenergy coupling of the scalar neutron flux at lattice level. This hitherto unavailable information does not entail a significant computational cost. In sodiumcooled fast reactor calculations, the quality of diffusion theory results can be improved by either using directional diffusion coefficients and a fine energy mesh, or via leakagecorrected 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 graphitemoderated, gascooled 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 2step procedure, with Monte Carlo as a fewgroup cross section data generator at lattice level, followed by deterministic multigroup 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 sodiumcooled fast reactor system, for which fewgroup cross section data was generated by stochastic methods with Serpent 2. The quality of the fullcore diffusion results is evaluated by contrasting system eigenvalues and power distributions against detailed, fullcore 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 designs. The newly proposed leakage model has a similar performance to that one of albedo iterations, and provides valuable information about the spaceenergy coupling of the scalar neutron flux at lattice level. This hitherto unavailable information does not entail a significant computational cost. In sodiumcooled fast reactor calculations, the quality of diffusion theory results can be improved by either using directional diffusion coefficients and a fine energy mesh, or via leakagecorrected 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 graphitemoderated, gascooled reactors.
KW  Monte Carlo
KW  Serpent
KW  directional diffusion coefficient
KW  TRIZ
KW  TRIVAC
KW  neutron leakage
KW  layer expansion
KW  albedo
KW  B1
KW  discontinuity factor
KW  SFR
M3  Dissertation
SN  9789526067353
T3  Aalto University publication series: Doctoral Dissertations
PB  Aalto University
CY  Espoo
ER 