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.
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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",
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publisher = "VTT Technical Research Centre of Finland",
number = "156",
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}

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.