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Numerical modelling of condensation pool (NUMPOOL). NUMPOOL summary report
Jarto Niemi
SAFIR2010. The Finnish Research Programme on Nuclear Power Plant Safety 2007-2010. Interim Report. Eija Karita Puska (Ed.). VTT Tiedotteita - Research Notes 2466, 160 - 172 Numerical methods for analyzing pressure suppression pools in boiling water reactors are developed. Some of the first experiments performed with the PPOOLEX facility are analysed with computational fluid dynamics (CFD) and Fluid-Structure Interaction (FSI) simulations. PPOOLEX vessel is a scaled-down model of a boiling water reactor containment with pressurized drywell and wetwell compartments. The compartments are connected with a vent pipe, whose outlet was submerged in a water pool of the wetwell. In the first experiments, air was blown into the drywell of the pressurized PPOOLEX vessel. The experiments and modelling are continued with steam discharges. Simulation of an air discharge was performed by using the Fluent CFD code. The pressurization of the drywell, clearance of the vent pipe and pressurization of ...
Impact 2014 (IMPACT2014) and structural mechanics analyses of soft and hard impacts (SMASH)
2013 •
arja saarenheimo
OpenFOAM CFD-solver for nuclear safety related flow simulations (NUFOAM)
2013 •
Jarto Niemi
Software Process: Improvement and Practice
Qualification of safety-critical systems in TVO nuclear power plants
2007 •
Juha Halminen
Probabilistic risk assessment method development and applications (PRAMEA)
2017 •
Marja Liinasuo
Numerical Modeling of Condensation Pool (NUMPOOL)
2011 •
Jarto Niemi
CFD Modelling of NPP Horizontal and Vertical Steam Generators (SGEN). SGEN summary report
Jarto Niemi
SAFIR2010. The Finnish Research Programme on Safety 2007-2010. Final Report. Puska, Eija Karita & Suolanen, Vesa. VTT Research Notes 2571, 225 - 236 A porosity model for the simulation of two-phase flow on the secondary side of both horizontal and vertical steam generators is presented. The Euler-Euler multiphase model of the Fluent CFD code is used for the simulation of the secondary side. The outer wall temperatures of the heat transfer tubes obtained by performing an APROS simulation of the steam generator are used for generating the source terms for the enthalpy equations. The model is tested by performing simulations for a horizontal steam generator of Loviisa VVER-440 nuclear power plant and a vertical steam generator of the PWR-PACTEL test facility. A porosity model for the simulation of two-phase flow on the secondary side of both horizontal and vertical steam generators is presented. The Euler-Euler multiphase model of the Fluent CFD code is used for the simulation of the s...
SAFIR2010. The Finnish Research Programme on Nuclear Power Plant Safety 2007-2010. Final Report
Teemu Reiman
VTT Tiedotteita - Research Notes 2571 Major part of Finnish public research on nuclear power plant safety during the years 2007-2010 has been carried out in the SAFIR2010 programme. The steering group of SAFIR2010 consisted of representatives from Radiation and Nuclear Safety Authority (STUK), Ministry of Employment and the Economy (MEE), Technical Research Centre of Finland (VTT), Teollisuuden Voima Oyj (TVO), Fortum Power and Heat Oyj, Fortum Nuclear Services Oy (Fortum), Finnish Funding Agency for Technology and Innovation (Tekes), Aalto University School of Science and Technology (Aalto, former Helsinki University of Technology) and Lappeenranta University of Technology (LUT). In addition to representatives of these organisations, the Steering Group had permanent experts from the Swedish Radiation Safety Authority (SSM) and Fennovoima Oy (Fennovoima). SAFIR2010 research programme was divided in eight research areas that were Organisation and human, Automation and control room, F...
Influence of Material, Environment and Strain Rate on Environmentally Assisted Cracking of Austenitic Nuclear Materials (DEFSPEED). DEFSPEED summary report
Mykola Ivanchenko
SAFIR2010. The Finnish Research Programme on Safety 2007-2010. Final Report. Puska, Eija Karita & Suolanen, Vesa. VTT Research Notes 2571, 440 - 452 The DEFSPEED project (Influence of material, environment and strain rate on environmentally assisted cracking on austenitic nuclear materials) focussed on increasing the understanding of environmentally assisted cracking (EAC) and irradiation assisted stress corrosion cracking (IASCC) mechanisms in austenitic nuclear materials. Several sub-tasks were performed to achieve the set objectives during the four year project. The Super Slow Strain Rate Testing technique (SSSRT) was developed and employed for initiation studies on austenitic nuclear materials in LWR environments followed by in-depth investigations on deformation distribution. Austenitic nuclear materials show dynamic strain ageing behaviour, as described in a thesis work. The effect of environment on fracture toughness properties in nickel-based welds at slow strain rate in hyd...
INTELI summary report
2004 •
arja saarenheimo
