General description of the PACTEL test facility

Jari Tuunanen, Jyrki Kouhia, Heikki Purhonen, Vesa Riikonen, Markku Puustinen, Scott Semken, Harri Partanen, Ilkka Saure, Hannu Pylkkö

Research output: Book/ReportReport


The PACTEL is a test facility designed to model the thermal-hydraulic behaviours of the Soviet-designed VVER-440 pressurized water reactors currently in use in Finland. These reactors have unique features that differ from other PWR designs. The PACTEL simulates the major components and systems of the reference PWR, making it possible to examine postulated small- and medium-break LOCA's and operational transients. The PACTEL is a volume-scaled model (1:305). To ensure that gravitational forces remain equal to those in the reference reactor, the major components and systems in the PACTEL preserve a 1:1 elevation equivalence to the reference reactor. Preserving the elevation equivalence and scaling by volume results in relatively small hydraulic diameters. This report describes some of the affects of the smaller hydraulic diameters on the thermal-hydraulic characteristics of the PACTEL. The PACTEL steam generator tube diameters and the tubes' angle-of-inclination are the same as in the reference reactor. Primary-side volume scaling while preserving elevation equivalence results in shorter tubes with double the vertical spacing. This results in oversized secondary-side volumes. This report also discusses the thermal-hydraulic affects of these larger secondary-side volumes. Assessing thermal-hydraulic computer codes used for the safety analyses of nuclear power plants is the final goal of the PACTEL experiment programmes. This report supplies the physical data on the PACTEL that analysts need to prepare their code models. It describes the PACTEL instrumentation and data acquisition system. The report also lists the experiments conducted prior to publishing and summarises each experiment procedure. Used in parallel with the Experiment Data Report, this PACTEL description provides all the information needed to analyse each experiment with a modern thermal-hydraulic computer code simulation.
Original languageEnglish
Place of PublicationEspoo
PublisherVTT Technical Research Centre of Finland
Number of pages110
ISBN (Electronic)951-38-5339-X
ISBN (Print)951-38-5338-1
Publication statusPublished - 1998
MoE publication typeD4 Published development or research report or study

Publication series

SeriesVTT Tiedotteita - Meddelanden - Research Notes


  • nuclear power plants
  • nuclear reactors
  • pressurized water reactors
  • test facilities
  • pressure vessels
  • pressure measurement
  • thermal-hydraulic performance


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