Studies on the assessment and validation of reactor dynamics models used in Finland: Dissertation

Timo Vanttola

    Research output: ThesisDissertationCollection of Articles

    Abstract

    Two reactor dynamics related computer codes of the calculation system at the Technical Research Centre of Finland have been assessed. The codes, TRAB and SMATRA, have been examined from two points of view: First, models of some critical phenomena determining the worst fuel rod conditions during reactor transients have been evaluated on the basis of experimental information. Second, the overall behaviour of the codes describing the dynamics of the reactor core and its cooling system has been studied on the basis of simulation of real transients and of performed safety analyses of selected accidents. The emphasis is on the VVER-440 reactors, but the generality of the methods has been demonstrated by showing that the key phenomena of the Chernobyl accident can be reproduced and analysed using the same calculation system. In this study the separate phenomena examined are single- and two-phase friction, post DNB heat transfer and critical heat flux in the VVER rod bundle. It appeared that the models of the pressure loss terms in the codes predict well the observed behaviour in a VVER specific test bundle. Well-known critical heat flux correlations have been tested against four data sets, and one of the correlations was modified to adapt the critical heat flux table presented by Kirillov. Unlike most other critical heat flux correlations, the models of Gidropress and, in particular, Smolin proved to cover a very wide parameter range with predictions mostly on the conservative side. These two correlations are used in the studied codes. On the other hand, some uncertainty must be associated with the modelling of the post dryout heat transfer, although the model used in the codes is probably on the conservative side in the VVER rod bundle. The SMATRA code reliably reproduced the four transients chosen as validation cases for the Loviisa VVER-440 nuclear power plant. The minor deviations observed in the simulations may, at least partly, be associated with uncertainties in the plant data. In the quoted safety analyses the axially one-dimensional neutron kinetics model of the codes, appended with the transverse synthesis model, appeared to be valuable in the calculation of the reactivity initiated accidents and the anticipated transients without scram. Some boron dilution cases, which may be critical for the safety of the power plant, have been singled out for particular investigation in this study.
    Original languageEnglish
    QualificationDoctor Degree
    Awarding Institution
    • Helsinki University of Technology
    Supervisors/Advisors
    • Salomaa, Rainer, Supervisor, External person
    Award date5 Nov 1993
    Place of PublicationEspoo
    Publisher
    Print ISBNs951-38-4394-7
    Publication statusPublished - 1993
    MoE publication typeG5 Doctoral dissertation (article)

    Fingerprint

    Dynamic models
    Heat flux
    Accidents
    Heat transfer
    Reactor cores
    Cooling systems
    Nuclear power plants
    Dilution
    Boron
    Power plants
    Neutrons
    Friction
    Kinetics
    Uncertainty

    Keywords

    • reactors
    • dynamics
    • mathematical models
    • computer codes
    • validation
    • heat transfer
    • friction
    • critical heat flux
    • TRAB
    • SMATRA
    • WWER type reactors
    • WWER-440 reactor
    • safety analysis

    Cite this

    Vanttola, T. (1993). Studies on the assessment and validation of reactor dynamics models used in Finland: Dissertation. Espoo: VTT Technical Research Centre of Finland.
    Vanttola, Timo. / Studies on the assessment and validation of reactor dynamics models used in Finland : Dissertation. Espoo : VTT Technical Research Centre of Finland, 1993. 192 p.
    @phdthesis{b3fa8b2cfa3943aea05f46295374fd3c,
    title = "Studies on the assessment and validation of reactor dynamics models used in Finland: Dissertation",
    abstract = "Two reactor dynamics related computer codes of the calculation system at the Technical Research Centre of Finland have been assessed. The codes, TRAB and SMATRA, have been examined from two points of view: First, models of some critical phenomena determining the worst fuel rod conditions during reactor transients have been evaluated on the basis of experimental information. Second, the overall behaviour of the codes describing the dynamics of the reactor core and its cooling system has been studied on the basis of simulation of real transients and of performed safety analyses of selected accidents. The emphasis is on the VVER-440 reactors, but the generality of the methods has been demonstrated by showing that the key phenomena of the Chernobyl accident can be reproduced and analysed using the same calculation system. In this study the separate phenomena examined are single- and two-phase friction, post DNB heat transfer and critical heat flux in the VVER rod bundle. It appeared that the models of the pressure loss terms in the codes predict well the observed behaviour in a VVER specific test bundle. Well-known critical heat flux correlations have been tested against four data sets, and one of the correlations was modified to adapt the critical heat flux table presented by Kirillov. Unlike most other critical heat flux correlations, the models of Gidropress and, in particular, Smolin proved to cover a very wide parameter range with predictions mostly on the conservative side. These two correlations are used in the studied codes. On the other hand, some uncertainty must be associated with the modelling of the post dryout heat transfer, although the model used in the codes is probably on the conservative side in the VVER rod bundle. The SMATRA code reliably reproduced the four transients chosen as validation cases for the Loviisa VVER-440 nuclear power plant. The minor deviations observed in the simulations may, at least partly, be associated with uncertainties in the plant data. In the quoted safety analyses the axially one-dimensional neutron kinetics model of the codes, appended with the transverse synthesis model, appeared to be valuable in the calculation of the reactivity initiated accidents and the anticipated transients without scram. Some boron dilution cases, which may be critical for the safety of the power plant, have been singled out for particular investigation in this study.",
    keywords = "reactors, dynamics, mathematical models, computer codes, validation, heat transfer, friction, critical heat flux, TRAB, SMATRA, WWER type reactors, WWER-440 reactor, safety analysis",
    author = "Timo Vanttola",
    note = "Project code: YDI1000537",
    year = "1993",
    language = "English",
    isbn = "951-38-4394-7",
    series = "VTT Publications",
    publisher = "VTT Technical Research Centre of Finland",
    number = "156",
    address = "Finland",
    school = "Helsinki University of Technology",

    }

    Vanttola, T 1993, 'Studies on the assessment and validation of reactor dynamics models used in Finland: Dissertation', Doctor Degree, Helsinki University of Technology, Espoo.

    Studies on the assessment and validation of reactor dynamics models used in Finland : Dissertation. / Vanttola, Timo.

    Espoo : VTT Technical Research Centre of Finland, 1993. 192 p.

    Research output: ThesisDissertationCollection of Articles

    TY - THES

    T1 - Studies on the assessment and validation of reactor dynamics models used in Finland

    T2 - Dissertation

    AU - Vanttola, Timo

    N1 - Project code: YDI1000537

    PY - 1993

    Y1 - 1993

    N2 - Two reactor dynamics related computer codes of the calculation system at the Technical Research Centre of Finland have been assessed. The codes, TRAB and SMATRA, have been examined from two points of view: First, models of some critical phenomena determining the worst fuel rod conditions during reactor transients have been evaluated on the basis of experimental information. Second, the overall behaviour of the codes describing the dynamics of the reactor core and its cooling system has been studied on the basis of simulation of real transients and of performed safety analyses of selected accidents. The emphasis is on the VVER-440 reactors, but the generality of the methods has been demonstrated by showing that the key phenomena of the Chernobyl accident can be reproduced and analysed using the same calculation system. In this study the separate phenomena examined are single- and two-phase friction, post DNB heat transfer and critical heat flux in the VVER rod bundle. It appeared that the models of the pressure loss terms in the codes predict well the observed behaviour in a VVER specific test bundle. Well-known critical heat flux correlations have been tested against four data sets, and one of the correlations was modified to adapt the critical heat flux table presented by Kirillov. Unlike most other critical heat flux correlations, the models of Gidropress and, in particular, Smolin proved to cover a very wide parameter range with predictions mostly on the conservative side. These two correlations are used in the studied codes. On the other hand, some uncertainty must be associated with the modelling of the post dryout heat transfer, although the model used in the codes is probably on the conservative side in the VVER rod bundle. The SMATRA code reliably reproduced the four transients chosen as validation cases for the Loviisa VVER-440 nuclear power plant. The minor deviations observed in the simulations may, at least partly, be associated with uncertainties in the plant data. In the quoted safety analyses the axially one-dimensional neutron kinetics model of the codes, appended with the transverse synthesis model, appeared to be valuable in the calculation of the reactivity initiated accidents and the anticipated transients without scram. Some boron dilution cases, which may be critical for the safety of the power plant, have been singled out for particular investigation in this study.

    AB - Two reactor dynamics related computer codes of the calculation system at the Technical Research Centre of Finland have been assessed. The codes, TRAB and SMATRA, have been examined from two points of view: First, models of some critical phenomena determining the worst fuel rod conditions during reactor transients have been evaluated on the basis of experimental information. Second, the overall behaviour of the codes describing the dynamics of the reactor core and its cooling system has been studied on the basis of simulation of real transients and of performed safety analyses of selected accidents. The emphasis is on the VVER-440 reactors, but the generality of the methods has been demonstrated by showing that the key phenomena of the Chernobyl accident can be reproduced and analysed using the same calculation system. In this study the separate phenomena examined are single- and two-phase friction, post DNB heat transfer and critical heat flux in the VVER rod bundle. It appeared that the models of the pressure loss terms in the codes predict well the observed behaviour in a VVER specific test bundle. Well-known critical heat flux correlations have been tested against four data sets, and one of the correlations was modified to adapt the critical heat flux table presented by Kirillov. Unlike most other critical heat flux correlations, the models of Gidropress and, in particular, Smolin proved to cover a very wide parameter range with predictions mostly on the conservative side. These two correlations are used in the studied codes. On the other hand, some uncertainty must be associated with the modelling of the post dryout heat transfer, although the model used in the codes is probably on the conservative side in the VVER rod bundle. The SMATRA code reliably reproduced the four transients chosen as validation cases for the Loviisa VVER-440 nuclear power plant. The minor deviations observed in the simulations may, at least partly, be associated with uncertainties in the plant data. In the quoted safety analyses the axially one-dimensional neutron kinetics model of the codes, appended with the transverse synthesis model, appeared to be valuable in the calculation of the reactivity initiated accidents and the anticipated transients without scram. Some boron dilution cases, which may be critical for the safety of the power plant, have been singled out for particular investigation in this study.

    KW - reactors

    KW - dynamics

    KW - mathematical models

    KW - computer codes

    KW - validation

    KW - heat transfer

    KW - friction

    KW - critical heat flux

    KW - TRAB

    KW - SMATRA

    KW - WWER type reactors

    KW - WWER-440 reactor

    KW - safety analysis

    M3 - Dissertation

    SN - 951-38-4394-7

    T3 - VTT Publications

    PB - VTT Technical Research Centre of Finland

    CY - Espoo

    ER -

    Vanttola T. Studies on the assessment and validation of reactor dynamics models used in Finland: Dissertation. Espoo: VTT Technical Research Centre of Finland, 1993. 192 p.