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 = "Aalto University Publication Series: Doctoral Dissertations",
    publisher = "Aalto University",
    number = "66",
    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 - Aalto University Publication Series: Doctoral Dissertations

    PB - Aalto University

    ER -