Development and applications of multi-physics capabilities in a continuous energy Monte Carlo neutron transport code

Dissertation

Research output: ThesisDissertationCollection of Articles

Abstract

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 two-stage 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 full-core 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 two-stage 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 time-dependent 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.
Original languageEnglish
QualificationDoctor Degree
Awarding Institution
  • Aalto University
Supervisors/Advisors
  • Tuomisto, Filip, Supervisor, External person
  • Leppänen, Jaakko, Advisor
Award date19 May 2017
Publisher
Print ISBNs978-952-60-7377-4, 978-951-38-8530-4
Electronic ISBNs978-952-60-7376-7, 978-951-38-8529-8
Publication statusPublished - 2017
MoE publication typeG5 Doctoral dissertation (article)

Fingerprint

neutrons
physics
energy
reactor cores
nuclear reactors
temperature distribution
neutron distribution
temperature
theses
coolants
approximation
iteration
economics
rods
simulation
assembly
reactors
methodology
estimates
interactions

Keywords

  • neutron transport
  • Monte Carlo
  • multi-physics
  • coupled calculation
  • reactor analysis

Cite this

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title = "Development and applications of multi-physics capabilities in a continuous energy Monte Carlo neutron transport code: Dissertation",
abstract = "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 two-stage 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 full-core 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 two-stage 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 time-dependent 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.",
keywords = "neutron transport, Monte Carlo, multi-physics, coupled calculation, reactor analysis",
author = "Ville Valtavirta",
year = "2017",
language = "English",
isbn = "978-952-60-7377-4",
series = "VTT Science",
publisher = "Aalto University",
number = "150",
address = "Finland",
school = "Aalto University",

}

Development and applications of multi-physics capabilities in a continuous energy Monte Carlo neutron transport code : Dissertation. / Valtavirta, Ville.

Aalto University, 2017. 172 p.

Research output: ThesisDissertationCollection of Articles

TY - THES

T1 - Development and applications of multi-physics 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 two-stage 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 full-core 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 two-stage 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 time-dependent 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 two-stage 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 full-core 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 two-stage 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 time-dependent 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 - multi-physics

KW - coupled calculation

KW - reactor analysis

M3 - Dissertation

SN - 978-952-60-7377-4

SN - 978-951-38-8530-4

T3 - VTT Science

PB - Aalto University

ER -