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/ReportReportProfessional

Abstract

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 typeNot Eligible

Publication series

NameVTT Tiedotteita - Meddelanden - Research Notes
PublisherVTT
No.1929
ISSN (Print)1235-0605
ISSN (Electronic)1455-0865

Fingerprint

hydraulics
experiment
nuclear power plant
test
data acquisition
instrumentation
spacing
reactor
safety
simulation
code
water

Keywords

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

Cite this

Tuunanen, J., Kouhia, J., Purhonen, H., Riikonen, V., Puustinen, M., Semken, S., ... Pylkkö, H. (1998). General description of the PACTEL test facility. Espoo: VTT Technical Research Centre of Finland. VTT Tiedotteita - Meddelanden - Research Notes, No. 1929
Tuunanen, Jari ; Kouhia, Jyrki ; Purhonen, Heikki ; Riikonen, Vesa ; Puustinen, Markku ; Semken, Scott ; Partanen, Harri ; Saure, Ilkka ; Pylkkö, Hannu. / General description of the PACTEL test facility. Espoo : VTT Technical Research Centre of Finland, 1998. 110 p. (VTT Tiedotteita - Meddelanden - Research Notes; No. 1929).
@book{cc2e7e2ed7894427b0fa51d0fb891071,
title = "General description of the PACTEL test facility",
abstract = "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.",
keywords = "nuclear power plants, nuclear reactors, pressurized water reactors, test facilities, pressure vessels, pressure measurement, thermal-hydraulic performance",
author = "Jari Tuunanen and Jyrki Kouhia and Heikki Purhonen and Vesa Riikonen and Markku Puustinen and Scott Semken and Harri Partanen and Ilkka Saure and Hannu Pylkk{\"o}",
note = "Project code: N6SU00046",
year = "1998",
language = "English",
isbn = "951-38-5338-1",
series = "VTT Tiedotteita - Meddelanden - Research Notes",
publisher = "VTT Technical Research Centre of Finland",
number = "1929",
address = "Finland",

}

Tuunanen, J, Kouhia, J, Purhonen, H, Riikonen, V, Puustinen, M, Semken, S, Partanen, H, Saure, I & Pylkkö, H 1998, General description of the PACTEL test facility. VTT Tiedotteita - Meddelanden - Research Notes, no. 1929, VTT Technical Research Centre of Finland, Espoo.

General description of the PACTEL test facility. / Tuunanen, Jari; Kouhia, Jyrki; Purhonen, Heikki; Riikonen, Vesa; Puustinen, Markku; Semken, Scott; Partanen, Harri; Saure, Ilkka; Pylkkö, Hannu.

Espoo : VTT Technical Research Centre of Finland, 1998. 110 p. (VTT Tiedotteita - Meddelanden - Research Notes; No. 1929).

Research output: Book/ReportReportProfessional

TY - BOOK

T1 - General description of the PACTEL test facility

AU - Tuunanen, Jari

AU - Kouhia, Jyrki

AU - Purhonen, Heikki

AU - Riikonen, Vesa

AU - Puustinen, Markku

AU - Semken, Scott

AU - Partanen, Harri

AU - Saure, Ilkka

AU - Pylkkö, Hannu

N1 - Project code: N6SU00046

PY - 1998

Y1 - 1998

N2 - 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.

AB - 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.

KW - nuclear power plants

KW - nuclear reactors

KW - pressurized water reactors

KW - test facilities

KW - pressure vessels

KW - pressure measurement

KW - thermal-hydraulic performance

M3 - Report

SN - 951-38-5338-1

T3 - VTT Tiedotteita - Meddelanden - Research Notes

BT - General description of the PACTEL test facility

PB - VTT Technical Research Centre of Finland

CY - Espoo

ER -

Tuunanen J, Kouhia J, Purhonen H, Riikonen V, Puustinen M, Semken S et al. General description of the PACTEL test facility. Espoo: VTT Technical Research Centre of Finland, 1998. 110 p. (VTT Tiedotteita - Meddelanden - Research Notes; No. 1929).