Quantification of input uncertainties based on veera reflooding experiments

Torsti Alku

Research output: Chapter in Book/Report/Conference proceedingConference article in proceedingsScientificpeer-review

1 Citation (Scopus)

Abstract

Propagation of input uncertainties through code runs and statistical analysis of the results has come to be one of the two prevailing methods when performing Best-Estimate Plus Uncertainty (BEPU) analyses with thermal-hydraulic system codes. To use the method, the uncertainties of the inputs need to be defined. A difficult task in this respect is the quantification of the uncertainties of the constitutive equations of the code. For this reason several methods to estimate the uncertainties of reflooding related correlations have been under review in the OECD/NEA-endorsed PREMIUM benchmark, in which VTT has also been taking part. In this paper the same methodologies that were utilized by VTT in PREMIUM for quantification of the uncertainties, the FFTBM and CIRCE methods, have been used with data from the VEERA reflooding experiments. The experiments were performed in the 1990's at the Lappeenranta University of Technology VEERA facility in Lappeenranta, Finland in partnership with VTT. The code used to perform the analyses was the APROS thermal-hydraulic system code, which has been developed by VTT in partnership with Fortum Ltd since 1986. The quantification results obtained with the VEERA data were furthermore compared to the VTT results obtained with the FEBA facility reflooding data during the PREMIUM benchmark. Some issues with the VEERA data were identified during the analysis, due to which a part of the data was excluded. Even then it was possible to successfully perform a process of quantification of the uncertainties with the methodologies involved. The comparisons with the PREMIUM results revealed that even with the issues with the data intriguing results could be reached.
Original languageEnglish
Title of host publication16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16)
PublisherAmerican Nuclear Society ANS
Pages3644-3657
Editioncd-rom
ISBN (Print)978-151081184-3
Publication statusPublished - 2015
MoE publication typeA4 Article in a conference publication
Event16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-16 - Chicago, United States
Duration: 30 Aug 20154 Sep 2015
Conference number: 16

Conference

Conference16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-16
Abbreviated titleNURETH-16
CountryUnited States
CityChicago
Period30/08/154/09/15

Fingerprint

Experiments
Hydraulics
Uncertainty
Constitutive equations
Statistical methods
Hot Temperature

Keywords

  • BEPU
  • circe
  • FFTBM
  • input uncertainty
  • physical models

Cite this

Alku, T. (2015). Quantification of input uncertainties based on veera reflooding experiments. In 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (cd-rom ed., pp. 3644-3657). American Nuclear Society ANS.
Alku, Torsti. / Quantification of input uncertainties based on veera reflooding experiments. 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16). cd-rom. ed. American Nuclear Society ANS, 2015. pp. 3644-3657
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Alku, T 2015, Quantification of input uncertainties based on veera reflooding experiments. in 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16). cd-rom edn, American Nuclear Society ANS, pp. 3644-3657, 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-16, Chicago, United States, 30/08/15.

Quantification of input uncertainties based on veera reflooding experiments. / Alku, Torsti.

16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16). cd-rom. ed. American Nuclear Society ANS, 2015. p. 3644-3657.

Research output: Chapter in Book/Report/Conference proceedingConference article in proceedingsScientificpeer-review

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Alku T. Quantification of input uncertainties based on veera reflooding experiments. In 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16). cd-rom ed. American Nuclear Society ANS. 2015. p. 3644-3657