Influence of cross-section uncertainties on the calculated results from a thermo-hydraulics code

Bruno Miglierini, Tomasz Kozlowski

Research output: Contribution to journalArticlepeer-review

Abstract

This study deals with the influence of uncertainties in selected cross-section data in a neutronics code while their impact on the results from thermo-hydraulics code is investigated. Particular cross-section data were defined according to the Phenomenon Identification and Ranking Tables (PIRT) for a pressurized water reactor. Uncertainties were identified for the following data: β – delayed neutron fraction precursors and transport, absorption, ν-fission, κ-fission, down scattering for moderator temperature, moderator density and fuel temperature. As a neutronics code, PARCS in conjunction with thermo-hydraulics code TRACE was applied. Uncertainty analyses were carried out by the code DAKOTA where coupled PARCS/TRACE calculations were run 146 times in order to meet a two-sided 95/95 confidence interval. In the thermo-hydraulic code thermal power of the reactor core and inner temperature of the fuel rod were studied. This research shows that the most influential parameters associated with cross-section parametrization are moderator temperature and fuel temperature.

Original languageEnglish (US)
Article number107598
JournalAnnals of Nuclear Energy
Volume145
DOIs
StatePublished - Sep 15 2020
Externally publishedYes

Keywords

  • Coupled calculation
  • Cross-section uncertainty
  • DAKOTA
  • RIA accident
  • VVER-1000

ASJC Scopus subject areas

  • Nuclear Energy and Engineering

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