Development of tall-3D test matrix for apros code validation

Ignas Mickus, Kaspar Kööp, Marti Jeltsov, Dmitry Grishchenko, Pavel Kudinov, Jari Lappalainen

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


APROS code is a multifunctional process simulator which combines System Thermal-Hydraulic (STH) capabilities with 1D/3D reactor core neutronics and full automation system modeling. It is applied for various tasks throughout the complete power plant life cycle including R&D, process and control engineering, and operator training. Currently APROS is being developed for evaluation of Generation IV conceptual designs using Lead-Bismuth Eutectic (LBE) alloy coolant. TALL-3D facility has been built at KTH in order to provide validation data for standalone and coupled STH and Computational Fluid Dynamics (CFD) codes. The facility consists of sections with measured inlet and outlet conditions for separate effect and integral effect tests (SETs and IETs). The design is aimed at reducing experimental uncertainties and allowing full separation of code validation from model input calibration. In this paper we present the development of experimental TALL-3D test matrix for comprehensive validation of APROS code. First, the representative separate effect and integral system response quantities (SRQs) are defined. Second, sources of uncertainties are identified and code sensitivity analysis is carried out to quantify the effects of code input uncertainties on the code prediction. Based on these results the test matrixes for calibration and validation experiments are determined in order to minimize the code input uncertainties. The applied methodology and the results are discussed in detail.
Original languageEnglish
Title of host publication16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16)
PublisherAmerican Nuclear Society (ANS)
ISBN (Print)978-1-5108-1184-3
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


Conference16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-16
Abbreviated titleNURETH-16
CountryUnited States


  • generation IV
  • lead-bismuth eutectic
  • validation
  • dynamic process simulation

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