Abstract
Original language  English 

Qualification  Doctor Degree 
Awarding Institution 

Supervisors/Advisors 

Award date  19 May 2017 
Publisher  
Print ISBNs  9789526073774, 9789513885304 
Electronic ISBNs  9789526073767, 9789513885298 
Publication status  Published  2017 
MoE publication type  G5 Doctoral dissertation (article) 
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Keywords
 neutron transport
 Monte Carlo
 multiphysics
 coupled calculation
 reactor analysis
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Development and applications of multiphysics capabilities in a continuous energy Monte Carlo neutron transport code : Dissertation. / Valtavirta, Ville.
Aalto University, 2017. 172 p.Research output: Thesis › Dissertation › Collection of Articles
TY  THES
T1  Development and applications of multiphysics capabilities in a continuous energy Monte Carlo neutron transport code
T2  Dissertation
AU  Valtavirta, Ville
PY  2017
Y1  2017
N2  The accurate modeling of nuclear reactors is essential to the safe and economic operation of current and future reactor types. Due to the physical feedback effects between the neutron distribution in the reactor core and the fuel and coolant material temperatures and densities, all of the three fields need to be solved simultaneously in order to obtain a solution for the behavior of the nuclear reactor. This coupled problem has traditionally been solved using a twostage approach for the neutron transport. In this approach, the neutron interaction properties of different parts of the reactor core are first averaged using accurate neutron transport methods in a series of small scale simulations. These averaged quantities are then used in a simplified neutron transport model to obtain the fullcore solution in a reasonable time allowing for multiple iterations between the solvers of the different physical fields. In recent years, advances in methodology as well as in computational power have made it possible to apply the accurate neutron transport methods directly to the full core problem, enabling modeling of the important feedback effects in more detail than has been possible with the twostage approach. Moving to direct modeling of the coupled problem with the accurate neutron transport methods initially developed for the lattice calculations in the first part of the twostage approach requires various changes to the neutron transport methods. In this thesis, capabilities required to solve the neutron transport part of the coupled problem are developed and implemented in the continuous energy Monte Carlo neutron transport code Serpent 2. Coupled calculation schemes were developed and implemented both for internally and externally coupled calculations for three different simulation types: Steady state calculations, burnup calculations and timedependent transient calculations. The new coupled calculation capabilities were applied to the effective fuel temperature approximation, in which the complex fuel temperature distribution in a fuel rod or a fuel assembly is replaced with a single effective temperature for the neutron transport calculations. The new capabilities made it possible to estimate the effects of this approximation by providing an accurate reference solution using realistic temperature distributions provided by an internally or externally coupled fuel temperature solver.
AB  The accurate modeling of nuclear reactors is essential to the safe and economic operation of current and future reactor types. Due to the physical feedback effects between the neutron distribution in the reactor core and the fuel and coolant material temperatures and densities, all of the three fields need to be solved simultaneously in order to obtain a solution for the behavior of the nuclear reactor. This coupled problem has traditionally been solved using a twostage approach for the neutron transport. In this approach, the neutron interaction properties of different parts of the reactor core are first averaged using accurate neutron transport methods in a series of small scale simulations. These averaged quantities are then used in a simplified neutron transport model to obtain the fullcore solution in a reasonable time allowing for multiple iterations between the solvers of the different physical fields. In recent years, advances in methodology as well as in computational power have made it possible to apply the accurate neutron transport methods directly to the full core problem, enabling modeling of the important feedback effects in more detail than has been possible with the twostage approach. Moving to direct modeling of the coupled problem with the accurate neutron transport methods initially developed for the lattice calculations in the first part of the twostage approach requires various changes to the neutron transport methods. In this thesis, capabilities required to solve the neutron transport part of the coupled problem are developed and implemented in the continuous energy Monte Carlo neutron transport code Serpent 2. Coupled calculation schemes were developed and implemented both for internally and externally coupled calculations for three different simulation types: Steady state calculations, burnup calculations and timedependent transient calculations. The new coupled calculation capabilities were applied to the effective fuel temperature approximation, in which the complex fuel temperature distribution in a fuel rod or a fuel assembly is replaced with a single effective temperature for the neutron transport calculations. The new capabilities made it possible to estimate the effects of this approximation by providing an accurate reference solution using realistic temperature distributions provided by an internally or externally coupled fuel temperature solver.
KW  neutron transport
KW  Monte Carlo
KW  multiphysics
KW  coupled calculation
KW  reactor analysis
M3  Dissertation
SN  9789526073774
SN  9789513885304
T3  Aalto University Publication Series: Doctoral Dissertations
PB  Aalto University
ER 