Two-phase flow measurements with spacer grid in a rod bundle geometry

Heat transfer effectiveness in nuclear rod bundles is of great importance to nuclear reactor safety and economics. An important design parameter is the Critical Heat Flux (CHF), which limits the transferred heat from the fuel to the coolant. The CHF is determined by flow behaviour, especially the turbulence created inside the fuel rod bundle. Adiabatic experiments can be used to characterize the flow behaviour separately from the heat transfer phenomena in diabatic flow. To enhance the turbulence, mixing vanes are attached to spacer grids, which hold the rods in place. The vanes either make the flow swirl around a single sub-channel or induce cross-mixing between adjacent sub-channels. In adiabatic two-phase conditions an important phenomenon that can be investigated is the effect of the spacer on canceling the lift force, which collects the small bubbles to the rod surfaces leading to decreased CHF in diabatic conditions and thus limits the reactor power. Computational Fluid Dynamics (CFD) can be used to simulate the flow numerically and to test how different spacer configurations affect the flow. Experimental data is needed to validate and verify the used CFD models. Especially the modeling of turbulence is challenging even for single-phase flow inside the complex sub-channel geometry. In two-phase flow other factors such as bubble dynamics further complicate the modeling. To investigate the spacer grid effect on two-phase flow, and to provide further experimental data for CFD validation, a series of experiments was run on an adiabatic sub-channel flow loop using a duct-type spacer grid with different configurations. Utilizing the wire-mesh sensor technology, the facility gives high resolution experimental data in both time and space. The experimental results indicate that the duct-type spacer grid is less effective in canceling the lift force effect than the egg-crate type spacer tested earlier.

Kipsin kiteytymiseen vaikuttavat tekijät ammoniumsulfaattikiteyttämön lämmönvaihtimissa

Työssä pyrittiin selvittämään syitä kipsin saostumiseen ammoniumsulfaattikiteyttämön putkilämmönvaihtimien pinnalle ja miten epätoivottua saostumista voitaisiin estää. Lämmönvaihtimissa virtaa ammoniumsulfaattia, jossa on epäpuhtautena kalsiumia, joka saostuu pinnoille kalsiumsulfaattina. Kirjallisuusosassa tarkasteltiin kiteytymisen mekanismia ja kipsin kiteytymiseen vaikuttavia tekijöitä. Saostumien estoaineita ja niiden vaikutusta kipsin kiteytymiseen sekä kipsin liukoisuutta ammoniumsulfaattiliuoksessa käsiteltiin myös kirjallisuusosassa. Kipsin kiteytymiseen vaikuttavia tekijöitä selvitettiin laboratoriokokeilla, joissa pyrittiin simuloimaan lämmönvaihdinta lämpövastuksella. Laboratoriokokeissa kokeiltiin erilaisia saostuman estoaineita ja pyrittiin löytämään prosessiin mahdollisimman tehokas kipsin kiteytymisen estoaine. Lämmönvaihtimien toiminnan tehokkuutta eli muodostuneen saostuman vaikutusta lämmönsiirtymiseen tutkittiin veden luovuttaman lämpövirran avulla. Lämmönvaihtimien tukkeutumista selvitettiin putkien vaihdon tarpeen perusteella. Kalsiumpitoisuuden vaihteluja prosessivirroissa selvitettiin kalsiumtaseen avulla. Saostumiseen vaikuttavien tekijöiden lisäksi selvitettiin mistä ja kuinka paljon kalsiumia kulkeutuu prosessiin ja poistuu sieltä. Työn tarkoituksena oli löytää ratkaisu, jolla epätoivottua saostumista lämmönvaihdin-ten pinnoille pystyttäisiin vähentämään joko kemiallisesti tai muuttamalla prosessi-muuttujia. Kalsiumia havaittiin olevan eniten pelkistämön sisäisissä ammoniumsul-faattiliuoskierroissa. Kalsiumtaseen perusteella kalsiumia poistuu pelkistämöltä eniten kipsinä ammoniumsulfaattituotteen mukana. Laboratoriokokeissa havaittiin polykar-boksylaattien estävän kipsin kasvua parhaiten, joskin estoaineen oikealla annostuksel-la havaittiin olevan suuri vaikutus. Lämmönvaihtimien saostuman havaittiin olevan kipsin ja glauberiitin seos. Vaippapuolen luovuttamien lämpövirran arvojen perusteel-la pystyttiin seuraamaan putkien tukkeutumista.

OL1/OL2-suojarakennuksen jäähdytteenmenetysonnettomuusanalyysit APROS 5.09 -ohjelmistolla

Tässä työssä on tutkittu OL1/OL2-suojarakennuksen käyttäytymistä jäähdytteenmenetysonnettomuuden eli LOCA:n aikana. Onnettomuuden simulointiin on kehitetty suojarakennusmalli suomalaiseen APROS 5.09 - ohjelmistoon sisältyvällä LP-koodilla (Lumped Parameter Code). Työssä on keskitytty suojarakennuksen kannalta oleellisimpien suureiden: kaasutilavuuksien paineen sekä lämpötilan ja lauhdutusaltaan lämpötilan ja pinnankorkeuden ajalliseen käyttäytymiseen. Mallinnetut LOCA:t ovat päähöyrylinjan ja sammutetun reaktorin jäähdytysjärjestelmän putkikatkoksia. Simulointeja on tehty laitoksen täyden tehon ja kuumavalmiuden lähtötiloissa ja tarkasteltavien suureiden käyttäytymistä on tutkittu erikseen valituilla konservatiivisilla oletuksilla. Laskentatapaukset rajoittuvat lyhyeen aikaväliin 27.78 tuntia kuvitellusta putkirikosta eteenpäin. Tuloksia on verrattu OL1/OL2- laitostoimittajan, Westinghouse Electric Sweden AB:n, tekemiin lisensiointianalyyseihin. Työssä on myös kuvattu APROS-laskennan epävarmuustekijöitä. Suojarakennuksen on todettu käyttäytyvän fysikaalisesti yhtenevästi lisensiointianalyyseissä ja APROS:lla tehdyissä vertailulaskuissa.

Prosessiautomaatio pinnankorkeusmittauksen koelaitteistoissa

Pinnankorkeuden tunteminen kiehutusvesireaktorin painesäiliössä on erittäin tärkeää sen turvallisuusvaikutusten takia. Pinnankorkeutta mitataan vesipatsaiden korkeutta havaitsevien paine-eromittausten avulla. Säteilyturvakeskuksen YVL-ohjeiden mukaan turvallisuuteen vaikuttavien mittausten täytyy noudattaa moninkertaistus- ja erilaisuusperiaatteita. Yleensä erilaisuusperiaatetta on toteutettu käyttämällä erityyppisiä paine-eromittareita, mutta erilaisella fysikaalisella toimintaperiaatteella oleva mittaus olisi parempi ja toteuttaisi paremmin erilaisuusperiaatetta. Uimurikytkin olisi tällainen fysikaalisesti eri periaatteeseen perustuva pinnankorkeuden mittauslaite. Ydinvoimalaan tarkoitettu teknologia tulee kelpoistaa riippumattoman tahon toimesta ennen käyttöönottoa. Kelpoistamiskokeita varten Lappeenrannan teknillisen yliopiston Ydinturvallisuuden tutkimusyksikköön rakennettiin vuosina 2011–2013 kaksi koelaitteistoa. Näillä koelaitteistoilla tutkittiin uimurikytkimien toimintaa ja ominaisuuksia erilaisissa kiehutusvesireaktorin käyttötilanteissa. Koelaitteistot tarvitsivat toimiakseen automaatiojärjestelmät, jotka suunniteltiin pääosin noudattamalla suunnittelun elinkaarimallia sekä automaatiosuunnittelun sisältökokonaisuuksia. Automaatiojärjestelmien suunnittelu aloitettiin määrittelemällä koejärjestelyjen asettamat vaatimukset, jonka jälkeen tehtiin teknologiavalinnat. Seuraavaksi suunniteltiin automaatiojärjestelmien logiikkaohjelmistot, joiden kuvaukseen tämä työ pääasiassa keskittyy. Logiikkaohjelmistot toteutettiin graafisella National Instruments LabView -ohjelmointikielellä. Logiikkaohjelmistojen tuli hoitaa tiedonkeruuta, käyttöautomaatiota, turvallisuustehtäviä sekä kokeisiin liittyviä erikoistehtäviä. Ohjelmistot saatiin esikokeiden aikana toimimaan halutusti, ja varsinaiset kokeet voitiin suorittaa ilman merkittäviä ongelmia.

Analysis of PACTEL small break LOCA experiment using TRACE

A small break loss-of-coolant accident (SBLOCA) is one of problems investigated in an NPP operation. Such accident can be analyzed using an experiment facility and TRACE thermal-hydraulic system code. A series of SBLOCA experiments was carried out on Parallel Channel Test Loop (PACTEL) facility, exploited together with Technical Research Centre of Finland VTT Energy and Lappeenranta University of Technology (LUT), in order to investigate two-phase phenomena related to a VVER-type reactor. The experiments and a TRACE model of the PACTEL facility are described in the paper. In addition, there is the TRACE code description with main field equations. At the work, calculations of a SBLOCA series are implemented and after the calculations, the thesis discusses the validation of TRACE and concludes with an assessment of the usefulness and accuracy of the code in calculating small breaks.

Experimental thermal hydraulic studies on the enhancement of safety of LWRs

The safe use of nuclear power plants (NPPs) requires a deep understanding of the functioning of physical processes and systems involved. Studies on thermal hydraulics have been carried out in various separate effects and integral test facilities at Lappeenranta University of Technology (LUT) either to ensure the functioning of safety systems of light water reactors (LWR) or to produce validation data for the computer codes used in safety analyses of NPPs. Several examples of safety studies on thermal hydraulics of the nuclear power plants are discussed. Studies are related to the physical phenomena existing in different processes in NPPs, such as rewetting of the fuel rods, emergency core cooling (ECC), natural circulation, small break loss-of-coolant accidents (SBLOCA), non-condensable gas release and transport, and passive safety systems. Studies on both VVER and advanced light water reactor (ALWR) systems are included. The set of cases include separate effects tests for understanding and modeling a single physical phenomenon, separate effects tests to study the behavior of a NPP component or a single system, and integral tests to study the behavior of the whole system. In the studies following steps can be found, not necessarily in the same study. Experimental studies as such have provided solutions to existing design problems. Experimental data have been created to validate a single model in a computer code. Validated models are used in various transient analyses of scaled facilities or NPPs. Integral test data are used to validate the computer codes as whole, to see how the implemented models work together in a code. In the final stage test results from the facilities are transferred to the NPP scale using computer codes. Some of the experiments have confirmed the expected behavior of the system or procedure to be studied; in some experiments there have been certain unexpected phenomena that have caused changes to the original design to avoid the recognized problems. This is the main motivation for experimental studies on thermal hydraulics of the NPP safety systems. Naturally the behavior of the new system designs have to be checked with experiments, but also the existing designs, if they are applied in the conditions that differ from what they were originally designed for. New procedures for existing reactors and new safety related systems have been developed for new nuclear power plant concepts. New experiments have been continuously needed.

Thermal-Hydraulic Studies on the Safety Of VVER-440 Type Nuclear Power Plants

This thesis includes several thermal hydraulic analyses related to the Loviisa WER 440 nuclear power plant units. The work consists of experimental studies, analysis of the experiments, analysis of some plant transits and development of a calculational model for calculation of boric acid concentrations in the reactor. In the first part of the thesis, in the case of won of boric acid solution behaviour during long term cooling period of LOCAs, experiments were performed in scaled down test facilities. The experimental data together with the results of RELAPS/MOD3 simulations were used to develop a model for calculations of boric acid concentrations in the reactor during LOCAs. The results of calculations showed that margins to critical concentrations that would lead to boric acid crystallization were large, both in the reactor core and in the lower plenum. This was mainly caused by the fact that water in the primary cooling circuit includes borax (Na)BsO,.IOHZO), which enters the reactor when ECC water is taken from the sump and greatly increases boric acid solubility in water. In the second part, in the case of simulation of horizontal steam generators, experiments were performed with PACTEL integral test loop to simulate loss of feedwater transients. The PACTEL experiments, as well as earlier REWET III natural circulation tests, were analyzed with RELAPS/MOD3 Version Sm5 code. The analysis showed that the code was capable of simulating the main events during the experiments. However, in the case of loss of secondary side feedwater the code was not completely capable to simulate steam superheating in the secondary side of the steam generators. The third part of the work consists of simulations of Loviisa VVER reactor pump trip transients with RELAPSlMODI Eur, RELAPS/MOD3 and CATHARE codes. All three codes were capable to simulate the two selected pump trip transients and no significant differences were found between the results of different codes. Comparison of the calculated results with the data measured in the Loviisa plant also showed good agreement.

Catalytic direct synthesis of hydrogen peroxide in a novel microstructured reactor

The direct synthesis from hydrogen and oxygen is a green alternative for production of hydrogen peroxide. However, this process suffers from two challenges. Firstly, mixtures of hydrogen and oxygen are explosive over a wide range of concentrations (4-94% H2 in O2). Secondly, the catalytic reaction of hydrogen and oxygen involves several reaction pathways, many of them resulting in water production and therfore decreasing selectivity. The present work deals with these two challenges. The safety problem was dealed by employing a novel microstructured reactor. Selectivity of the reaction was highly improved by development a set of new catalysts. The final goal was to develop an effective and safe continuous process for direct synthesis of hydrogen peroxide from H2 and O2. Activated carbon cloth and Sibunit were examined as the catalysts’ supports. Palladium and gold monometallic and palladium-gold bimetallic catalysts were thoroughly investigated by numerous kinetic experiments performed in a tailored batch reactor and several catalyst charachterization methods. A complete set of data for direct synthesis of H2O2 and its catalytic decomposition and hydrogenation was obtained. These data were used to assess factors influencing selectivity and activity of the catalysts in direct synthesis of H2O2 as well as its decomposition and hydrogenation. A novel microstructured reactor was developed based on hydrodynamics and mass transfer studies in prototype microstractural plates. The shape and the size of the structural elements in the microreactor plate were optimized in a way to get high gas-liquid interfacial area and gas-liquid mass transfer. Finally, empirical correlations for the volumetric mass transfer coefficient were derived. A bench-scale continuous process was developed by using the novel microstructral plate reactor. A series of kinetic experiments were performed to investigate the effects of the gas and the liquid feed rates and their ratio, the amount of the catalyst, the gas feed composition and pressure on the final rate of H2O2 production and selectivity.

APROS code benchmarking inboiling water reactor accident analyses

Diplomityö käsittelee kiehutusvesilaitosten transienttien ja onnettomuuksien analysointia APROS-ohjelmiston avulla. Työ on tehty Teollisuuden Voima Oy:n (TVO) Olkiluoto 1 ja 2 laitosyksiköiden mallin pohjalta. Raportissa esitetään ohjelmiston käyttämiä yhtälöitäja laskentamalleja yleisellä tasolla. Työssä esitellään laitoksen yleispiirteet turvallisuustoimintoineen ja kuvataan ohjelmaan suureksi osaksi aiemmin luotua laskentamallia. Työssä on luetteloitu voimassa olevatlisensiointianalyysit, joiden joukosta on valittu laskentatapauksia ohjelmiston suorituskyvyn arviointia varten. Lisäksi työhön on valittu laskentatapauksia muilla kuin lisensointiin käytetyillä ohjelmilla lasketuista analyyseistä. Lisäksi on suoritettu vertailulaskuja konservatiivisen ja realistisen mallin erojen esille saamiseksi. Laskentatapauksia ovat mm. ylipainetransientti, jäähdytteen menetysonnettomuus ja oletettavissa oleva käyttöhäiriö, jossa pikasulku ei toimi (ATWS). Diplomityön edetessä laitosmallia on kehitetty edelleen lisäämällä joitakin järjestelmiä ja tarkentamalla joidenkin komponenttien kuvausta. Työssä ilmeni, että APROS soveltuu jäähdytteenmenetysonnettomuuden ja suojarakennuksen yhtäaikaiseen analyysiin. APROS.n vaste nopeisiin transientteihin jäi kuitenkin vertailutasosta. Tämän työn perusteella APROS-mallia kehitys jatkuu edelleen siten, että se soveltuisi entistä paremmin myös nopeiden transienttien ja ATWS-tilanteiden kuvaamiseen. Työssä olevaa lisensointianalyysien kuvausta tullaan käyttämään hyväksi selvitettäessä laitoksen turvallisuuden väliarviossa tarvittavien analyysien määrää ja laatua. Nyt saatuja kokemuksia voidaan hyödyntää myös mahdollisen kolmiulotteisen sydänmallin hankinnassa APROS-ohjelmistoon. Tässä diplomityössä esitettyjä parannuksia voidaan käyttää hyväksi SAFIRtutkimusohjelman hankkeiden suunnittelussa.

Validation Studies of Thermal-hydraulic Code for Safety Analysis of Nuclear Power Plants

This thesis gives an overview of the validation process for thermal hydraulic system codes and it presents in more detail the assessment and validation of the French code CATHARE for VVER calculations. Three assessment cases are presented: loop seal clearing, core reflooding and flow in a horizontal steam generator. The experience gained during these assessment and validation calculations has been used to analyze the behavior of the horizontal steam generator and the natural circulation in the geometry of the Loviisa nuclear power plant. The cases presented are not exhaustive, but they give a good overview of the work performed by the personnel of Lappeenranta University of Technology (LUT). Large part of the work has been performed in co-operation with the CATHARE-team in Grenoble, France. The design of a Russian type pressurized water reactor, VVER, differs from that of a Western-type PWR. Most of thermal-hydraulic system codes are validated only for the Western-type PWRs. Thus, the codes should be assessed and validated also for VVER design in order to establish any weaknesses in the models. This information is needed before codes can be used for the safety analysis. Theresults of the assessment and validation calculations presented here show that the CATHARE code can be used also for the thermal-hydraulic safety studies for VVER type plants. However, some areas have been indicated which need to be reassessed after further experimental data become available. These areas are mostly connected to the horizontal stem generators, like condensation and phase separation in primary side tubes. The work presented in this thesis covers a large numberof the phenomena included in the CSNI code validation matrices for small and intermediate leaks and for transients. Also some of the phenomena included in the matrix for large break LOCAs are covered. The matrices for code validation for VVER applications should be used when future experimental programs are planned for code validation.

Effect of Noncondensable Gases on Circulation of Primary Coolant in Nuclear Power Plants in Abnormal Situations

The present study focuses on two effects of the presence of a noncondensable gas on the thermal-hydraulic behavior of thecoolant of the primary circuit of a nuclear reactor in the VVER-440 geometry inabnormal situations. First, steam condensation with the presence of air was studied in the horizontal tubes of the steam generator (SG) of the PACTEL test facility. The French thermal-hydraulic CATHARE code was used to study the heat transfer between the primary and secondary side in conditions derived from preliminary experiments performed by VTT using PACTEL. In natural circulation and single-phase vapor conditions, the injection of a volume of air, equivalent to the totalvolume of the primary side of the SG at the entrance of the hot collector, did not stop the heat transfer from the primary to the secondary side. The calculated results indicate that air is located in the second half-length (from the mid-length of the tubes to the cold collector) in all the tubes of the steam generator The hot collector remained full of steam during the transient. Secondly, the potential release of the nitrogen gas dissolved in the water of the accumulators of the emergency core coolant system of the Loviisa nuclear power plant (NPP) was investigated. The author implemented a model of the dissolution and release ofnitrogen gas in the CATHARE code; the model created by the CATHARE developers. In collaboration with VTT, an analytical experiment was performed with some components of PACTEL to determine, in particular, the value of the release time constant of the nitrogen gas in the depressurization conditions representative of the small and intermediate break transients postulated for the Loviisa NPP. Such transients, with simplified operating procedures, were calculated using the modified CATHARE code for various values of the release time constant used in the dissolution and release model. For the small breaks, nitrogen gas is trapped in thecollectors of the SGs in rather large proportions. There, the levels oscillate until the actuation of the low-pressure injection pumps (LPIS) that refill the primary circuit. In the case of the intermediate breaks, most of the nitrogen gas is expelled at the break and almost no nitrogen gas is trapped in the SGs. In comparison with the cases calculated without taking into account the release of nitrogen gas, the start of the LPIS is delayed by between 1 and 1.75 h. Applicability of the obtained results to the real safety conditions must take into accountthe real operating procedures used in the nuclear power plant.

Design and implementation of a multipurpose reactor for scale-up studies

In the theoretical part, the different polymerisation catalysts are introduced and the phenomena related to mixing in the stirred tank reactor are presented. Also the advantages and challenges related to scale-up are discussed. The aim of the applied part was to design and implement an intermediate-sized reactor useful for scale-up studies. The reactor setting was tested making one batch of Ziegler–Natta polypropylene catalyst. The catalyst preparation with a designed equipment setting succeeded and the catalyst was analysed. The analyses of the catalyst were done, because the properties of the catalyst were compared to the normal properties of Ziegler–Natta polypropylene catalyst. The total titanium content of the catalyst was slightly higher than in normal Ziegler–Natta polypropylene catalyst, but the magnesium and aluminium content of the catalyst were in the normal level. By adjusting the siphonation tube and adding one washing step the titanium content of the catalyst could be decreased. The particle size of the catalyst was small, but the activity was in a normal range. The size of the catalyst particles could be increased by decreasing the stirring speed. During the test run, it was noticed that some improvements for the designed equipment setting could be done. For example more valves for the chemical feed line need to be added to ensure inert conditions during the catalyst preparation. Also nitrogen for the reactor needs to separate from other nitrogen line. With this change the pressure in the reactor can be kept as desired during the catalyst preparation. The proposals for improvements are presented in the applied part. After these improvements are done, the equipment setting is ready for start-up. The computational fluid dynamics model for the designed reactor was provided by cooperation with Lappeenranta University of Technology. The experiments showed that for adequate mixing with one impeller, stirring speed of 600 rpm is needed. The computational fluid dynamics model with two impellers showed that there was no difference in the mixing efficiency if the upper impeller were pumping downwards or upwards.

Simulation of hydrogen peroxide direct synthesis in a microreactor

A set of models in Aspen plus was built to simulate the direct synthesis process of hydrogen peroxide in a micro-reactor system. This process model can be used to carry out material balance calculation under various experimental conditions. Three thermodynamic property methods were compared by calculating gas solubility and Uniquac-RK method was finally selected for process model. Two different operation modes with corresponding operation conditions were proposed as the starting point of future experiments. Simulations for these two modes were carried out to get the information of material streams. Moreover, some hydrodynamic parameters such as gas/liquid superficial velocity, gas holdup were also calculated with improved process model. These parameters proved the proposed experimental conditions reasonable to some extent. The influence of operation conditions including temperature, pressure and circulation ratio was analyzed for the first operation mode, where pure oxygen was fed into dissolving tank and hydrogen-carbon dioxide mixture was fed into microreactor directly. The preferred operation conditions for the system are low temperature (2°C) and high pressure (30 bar) in dissolving tank. High circulation ratio might be good in the sense that more oxygen could be dissolved and fed into reactor for reactions, but meanwhile hydrodynamics of microreactor should be considered. Furthermore, more operation conditions of reactor gas/liquid feeds in both of two operation modes were proposed to provide guidance for future experiment design and corresponding hydrodynamic parameters were also calculated. Finally, safety issue was considered from thermodynamic point of view and there is no explosion danger at given experimental plan since the released reaction heat will not cause solvent vaporization inside the microchannels. The improvement of process model still needs further study based on the future experimental results.

APROS model validation in case of fast transients in a boiling water reactor

This Master´s thesis investigates the performance of the Olkiluoto 1 and 2 APROS model in case of fast transients. The thesis includes a general description of the Olkiluoto 1 and 2 nuclear power plants and of the most important safety systems. The theoretical background of the APROS code as well as the scope and the content of the Olkiluoto 1 and 2 APROS model are also described. The event sequences of the anticipated operation transients considered in the thesis are presented in detail as they will form the basis for the analysis of the APROS calculation results. The calculated fast operational transient situations comprise loss-of-load cases and two cases related to a inadvertent closure of one main steam isolation valve. As part of the thesis work, the inaccurate initial data values found in the original 1-D reactor core model were corrected. The input data needed for the creation of a more accurate 3-D core model were defined. The analysis of the APROS calculation results showed that while the main results were in good accordance with the measured plant data, also differences were detected. These differences were found to be caused by deficiencies and uncertainties related to the calculation model. According to the results the reactor core and the feedwater systems cause most of the differences between the calculated and measured values. Based on these findings, it will be possible to develop the APROS model further to make it a reliable and accurate tool for the analysis of the operational transients and possible plant modifications.

Utilization of Particle Image velocimetry in PPOOLEX condensation experiments

Particle Image Velocimetry, PIV, is an optical measuring technique to obtain velocity information of a flow in interest. With PIV it is possible to achieve two or three dimensional velocity vector fields from a measurement area instead of a single point in a flow. Measured flow can be either in liquid or in gas form. PIV is nowadays widely applied to flow field studies. The need for PIV is to obtain validation data for Computational Fluid Dynamics calculation programs that has been used to model blow down experiments in PPOOLEX test facility in the Lappeenranta University of Technology. In this thesis PIV and its theoretical background are presented. All the subsystems that can be considered to be part of a PIV system are presented as well with detail. Emphasis is also put to the mathematics behind the image evaluation. The work also included selection and successful testing of a PIV system, as well as the planning of the installation to the PPOOLEX facility. Already in the preliminary testing PIV was found to be good addition to the measuring equipment for Nuclear Safety Research Unit of LUT. The installation to PPOOLEX facility was successful even though there were many restrictions considering it. All parts of the PIV system worked and they were found out to be appropriate for the planned use. Results and observations presented in this thesis are a good background to further PIV use.