Fischer-Tropsch Slurry Bubble Column Reactor Modeling

The literature part of the work reviews overall Fischer-Tropsch process, Fischer-Tropsch reactors and catalysts. Fundamentals of Fischer-Tropsch modeling are also presented. The emphasis is on the reactor unit. Comparison of the reactors and the catalysts is carried out to choose the suitable reactor setup for the modeling work. The effects of the operation conditions are also investigated. Slurry bubble column reactor model operating with cobalt catalyst is developed by taking into account the mass transfer of the reacting components (CO and H2) and the consumption of the reactants in the liquid phase. The effect of hydrostatic pressure and the change in total mole flow rate in gas phase are taken into account in calculation of the solubilities. The hydrodynamics, reaction kinetics and product composition are determined according to literature. The cooling system and furthermore the required heat transfer area and number of cooling tubes are also determined. The model is implemented in Matlab software. Commercial scale reactor setup is modeled and the behavior of the model is investigated. The possible inaccuraries are evaluated and the suggestions for the future work are presented. The model is also integrated to Aspen Plus process simulation software, which enables the usage of the model in more extensive Fischer-Tropsch process simulations. Commercial scale reactor of diameter of 7 m and height of 30 m was modeled. The capacity of the reactor was calculated to be about 9 800 barrels/day with CO conversion of 75 %. The behavior of the model was realistic and results were in the right range. The highest uncertainty to model was estimated to be caused by the determination of the kinetic rate.

Coolant Flow and Heat Transfer in Pebble-Bed Reactor Core

This thesis gathers knowledge about ongoing high-temperature reactor projects around the world. Methods for calculating coolant flow and heat transfer inside a pebble-bed reactor core are also developed. The thesis begins with the introduction of high-temperature reactors including the current state of the technology. Process heat applications that could use the heat from a high-temperature reactor are also introduced. A suitable reactor design with data available in literature is selected for the calculation part of the thesis. Commercial computational fluid dynamics software Fluent is used for the calculations. The pebble-bed is approximated as a packed-bed, which causes sink terms to the momentum equations of the gas flowing through it. A position dependent value is used for the packing fraction. Two different models are used to calculate heat transfer. First a local thermal equilibrium is assumed between the gas and solid phases and a single energy equation is used. In the second approach, separate energy equations are used for the phases. Information about steady state flow behavior, pressure loss, and temperature distribution in the core is obtained as results of the calculations. The effect of inlet mass flow rate to pressure loss is also investigated. Data found in literature and the results correspond each other quite well, considered the amount of simplifications in the calculations. The models developed in this thesis can be used to solve coolant flow and heat transfer in a pebble-bed reactor, although additional development and model validation is needed for better accuracy and reliability.

Gamma spectrometry and gamma and X-ray tomography of nuclear fuel

The purpose of gamma spectrometry and gamma and X-ray tomography of nuclear fuel is to determine both radionuclide concentration and integrity and deformation of nuclear fuel. The aims of this thesis have been to find out the basics of gamma spectrometry and tomography of nuclear fuel, to find out the operational mechanisms of gamma spectrometry and tomography equipment of nuclear fuel, and to identify problems that relate to these measurement techniques. In gamma spectrometry of nuclear fuel the gamma-ray flux emitted from unstable isotopes is measured using high-resolution gamma-ray spectroscopy. The production of unstable isotopes correlates with various physical fuel parameters. In gamma emission tomography the gamma-ray spectrum of irradiated nuclear fuel is recorded for several projections. In X-ray transmission tomography of nuclear fuel a radiation source emits a beam and the intensity, attenuated by the nuclear fuel, is registered by the detectors placed opposite. When gamma emission or X-ray transmission measurements are combined with tomographic image reconstruction methods, it is possible to create sectional images of the interior of nuclear fuel. MODHERATO is a computer code that simulates the operation of radioscopic or tomographic devices and it is used to predict and optimise the performance of imaging systems. Related to the X-ray tomography, MODHERATO simulations have been performed by the author. Gamma spectrometry and gamma and X-ray tomography are promising non-destructive examination methods for understanding fuel behaviour under normal, transient and accident conditions.

Determining risk potential through accidents and near-accidents: an empirical study in a steel factory

Fatal and permanently disabling accidents form only one per I cent of all occupational accidents but in many branches of industry they account for more than half the accident costs. Furthermore the human suffering of the victim and his family is greater in severe accidents than in slight ones. For both human and economic reasons the severe accident risks should be identified befor injuries occur. It is for this purpose that different safety analysis methods have been developed . This study shows two new possible approaches to the problem.. The first is the hypothesis that it is possible to estimate the potential severity of accidents independent of the actual severity. The second is the hypothesis that when workers are also asked to report near accidents, they are particularly prone to report potentially severe near accidents on the basis of their own subjective risk assessment. A field study was carried out in a steel factory. The results supported both the hypotheses. The reliability and the validity of post incident estimates of an accident's potential severity were reasonable. About 10 % of accidents were estimated to be potentially critical; they could have led to death or very severe permanent disability. Reported near accidents were significantly more severe, about 60 $ of them were estimated to be critical. Furthermore the validity of workers subjective risk assessment, manifested in the near accident reports, proved to be reasonable. The studied new methods require further development and testing. They could be used both in routine usage in work places and in research for identifying and setting the priorities of accident risks.

Program hosting in communications environment with Reactor and thread pool approach

In this thesis concurrent communication event handling is implemented using thread pool approach. Concurrent events are handled with a Reactor design pattern and multithreading is implemented using a Leader/Followers design pattern. Main focus is to evaluate behaviour of implemented model by different numbers of concurrent connections and amount of used threads. Furthermore, model feasibility in a PeerHood middleware is evaluated. Implemented model is evaluated with created test environment which enables concurrent message sending from multiple connections to the system under test. Messages round trip times are measured in the tester application. In the evaluation processing delay into system is simulated and influence of delay to the average round trip time is analysed.

Passiiviset turvallisuusjärjestelmät tulevaisuuden kiehutusvesireaktorilaitoksissa

Kandidaatintyössä on esitelty passiivisten turvallisuusjärjestelmien hyödyntämistä seuraavan sukupolven kiehutusvesireaktorilaitoksissa.

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.

Reaktoreiden termisen tehon laskennan epävarmuuksien kartoitus

Diplomityössä selvitettiin Fortum Power and Heat Oy:n Loviisan VVER-440 painevesireaktorilaitosten termisen tehon laskentaan liittyviä epävarmuuksia. Laitoksen turvallisuusteknisissä käyttöehdoissa (TTKE) määrätään reaktorin suurimmaksi sallituksi lämpötehoksi 1500 MW. Tähän perustuen haluttiin selvittää nykyiseen RT1 laskentaan liittyvät epävarmuudet tarkastamalla nykyinen laskenta ja siinä käytetyt termohydrauliset laskentasovitteet. Työn alussa selostetaan lyhyesti Loviisan voimalaitoksen toimintaperiaate, jonka jälkeen esitellään laskentaan osallistuvat prosessimittaukset ja niihin liittyvät epävarmuustekijät. Mittauksille määritettiin epävarmuudet käyttäen hyödyksi komponenttivalmistajien tietoja sekä laitoksen kalibrointitodistuksia ja näiden lisäksi laskettiin standardin mukainen virhe virtauslaipoille. Edellä mainittujen virheiden perusteella voitiin laskea tehon epävarmuudet yksittäiselle höyrystimelle, josta edelleen varianssien summamenetelmällä saatiin reaktorin termiselle teholle 0,78 %:n epävarmuus 95 % luottamustasolla. Laskettua tehon epävarmuutta verrattiin Monte Carlo -menetelmällä suoritettuun tarkistuslaskentaan, jolla termisen tehon epävarmuudeksi saatiin 0,53 %, luottamustason ollessa 95 %. Työssä tarkasteltiin keskiarvotuksen vaikutusta mittausdataan. Näissä tarkasteluissa havaittiin pinnansäädöstä aiheutuva reaktoritehon huojunta, joka oli työn merkittävin havainto.

Quasi-objects and actors in the web of belief. Formation of a national nuclear waste management system: the case of Finland

Paper presented in ISA RC23 meeting, Gothenburg July 16th 2010

Introduktionen av kärnkraften i Finland

Finlands industri har av tradition varit starkt energikrävande. Träförädlingsindustrin, som fick sin egentliga start i medlet på 1800-talet, använde stora mängder energi liksom metallförädlingsföretagen i ett senare skede. Krigstiden med sin energiransonering visade handgripligen för allmänheten liksom för specialisterna att en tillräcklig tillgång till energi är ett livsvillkor för vår industri och därmed för vårt land. Efterkrigstiden kännetecknades av en allt snabbare utbyggnad av den på vatten- och ångkraft baserade elkraftskapaciteten, en utbyggnad som den inhemska verkstadsindustrin i stor utsträckning deltog i. Men redan på 1950-talet var vattenkraften till stor del utbyggd, varför den privata såväl som den statliga sektorns intresse allt mera inriktade sig på den speciellt i USA favoriserade atomenergin. Efter fördjupade studier i kärnfysik och kärnteknik vid the International School of Nuclear Science and Engineering i USA deltog författaren av dessa rader intensivt (först som Ahlströmanställd och senare som VD för Finnatom) i den utvecklingsverksamhet inom det kärntekniska området som inte bara elproducenterna utan även verkstadsindustrin i vårt land genomförde. Det var därför naturligt för mig att som objekt för min doktorsavhandling välja introduktionen av kärnkraften i Finland med speciell fokus på den inhemska verkstadsindustrins roll. Jag ställde följande forskningsfrågor: a. När och hur skedde introduktionen av kärnkraften i Finland? b. Vilka var orsakerna till och resultatet av denna introduktion? c. Vilken var den inhemska verkstadsindustrins roll? Ett grundligt studium av litteraturen inklusive mötesprotokoll och tidningsreferat samt personligen genomförda intervjuer med ett trettiotal av de verkliga aktörerna i den långa och komplicerade introduktionsprocessen ledde till en teori, vars riktighet jag anser mig ha kunnat bevisa. Den inhemska verkstadsindustrins roll var synnerligen central. Dess representanter lyckades, bl.a. refererande till erfarenheterna från utbyggnaden av vatten- och ångkraften liksom till byggandet av den underkritiska milan YXP samt forskningsreaktorn TRIGA, övertyga beslutsfattarna om att den besatt nödig kompetens för att kompensera den kompetensbrist som kunde iakttas inom vissa områden hos den sovjetiska kärnkraftverksleverantören. De inhemska leveranserna påverkade även driftsresultatet, speciellt i fallet Lovisa, i positiv riktning. Introduktionsprocessen, som omfattade tiden från slutet av 1950-talet till början på 1980-talet, beskrevs, noterande bl.a. J. W. Creswells anvisningar, i detalj i avhandlingen. Introduktionen fick som resultat konkurrenskraftig elkraft, impuls till start av nya företag, exempelvis Nokia Elektronik, liksom en klar höjning av den tekniska nivån hos vår industri, inkluderande kärnteknisk tillverkning i stor skala. Katastrofen i Tjernobyl i slutet av april 1986 innebar emellertid att utvecklingen tog en paus på ett par decennier. Erfarenheterna från introduktionsfasen kan förhoppningsvis utnyttjas till fullo nu, när utbyggnaden av kärnkraften återupptagits i vårt land.

Loviisan ydinvoimalaitoksen todennäköisyyspohjaisen riskianalyysin tason 2 epävarmuustarkastelut

Diplomityössä tarkastellaan Loviisan ydinvoimalaitoksen todennäköisyyspohjaisen riskianalyysin tason 2 epävarmuuksia. Tason 2 riskitutkimuksissa tutkitaan ydinvoimalaitosonnettomuuksia, joiden seurauksena osa reaktorin radioaktiivisista aineista vapautuu ympäristöön. Näiden tutkimuksien päätulos on suuren päästön vuotuinen taajuus ja se on pääosin todelliseen laitoshistoriaan perustuva tilastollinen odotusarvo. Tämän odotusarvon uskottavuutta voidaan parantaa huomioimalla merkittävimmät laskentaan liittyvät epävarmuudet. Epävarmuuksia laskentaan aiheutuu muiden muassa vakavan reaktorionnettomuuden ilmiöistä, turvallisuusjärjestelmien laitteista, inhimillisistä toiminnoista sekä luotettavuusmallin määrittelemättömistä osista. Diplomityössä kuvataan, kuinka epävarmuustarkastelut integroidaan osaksi Loviisan ydinvoimalaitoksen todennäköisyyspohjaisia riskianalyysejä. Tämä toteutetaan diplomityössä kehitetyillä apuohjelmilla PRALA:lla ja PRATU:lla, joiden avulla voidaan lisätä laitoshistorian perusteella muodostetut epävarmuusparametrit osaksi riskianalyysien luotettavuusdataa. Lisäksi diplomityössä on laskettu laskentaesimerkkinä Loviisan ydinvoimalaitoksen suuren päästön vuotuisen taajuuden vaihtelua kuvaava luottamusväli. Tämä laskentaesimerkki pohjautuu pääosin konservatiivisiin epävarmuusarvioihin, ei todellisiin tilastollisiin epävarmuuksiin. Laskentaesimerkin tulosten perusteella Loviisan suuren päästön taajuudella on laaja vaihteluväli; virhekertoimeksi saatiin 8,4 nykyisillä epävarmuusparametreilla. Suuren päästön taajuuden luottamusväliä voidaan kuitenkin tulevaisuudessa supistaa, kun hyödynnetään todelliseen laitoshistoriaan perustuvia epävarmuusparametreja.

Nuclear Matrix in Apoptotic Cell Death and Cell Proliferation. The role of Nuclear Mitotic Apparatus (NuMA) Protein

The nucleus is a membrane enclosed organelle containing most of the genetic information of the cell in the form of chromatin. The nucleus, which can be divided into many sub-organelles such as the nucleoli, the Cajal bodies and the nuclear lamina, is the site for several essential cellular functions such as the DNA replication and its regulation and most of the RNA synthesis and processing. The nucleus is often affected in disease: the size and the shape of the nucleus, the chromatin distribution and the size of the nucleoli have remained the basis for the grading of several cancers. The maintenance of the vertebrate body shape depends on the skeleton. Similarly, in a smaller context, the shape of the cell and the nucleus are mainly regulated by the cytoskeletal and nucleoskeletal elements. The nuclear matrix, which by definition is a detergent, DNase and salt resistant proteinaceous nuclear structure, has been suggested to form the nucleoskeleton responsible for the nuclear integrity. Nuclear mitotic apparatus protein, NuMA, a component of the nuclear matrix, is better known for its mitotic spindle organizing function. NuMA is one of the nuclear matrix proteins suggested to participate in the maintenance of the nuclear integrity during interphase but its interphase function has not been solved to date. This thesis study concentrated on the role of NuMA and the nuclear matrix as structural and functional components of the interphase nucleus. The first two studies clarified the essential role of caspase-3 in the disintegration of the nuclear structures during apoptosis. The second study also showed NuMA and chromatin to co-elute from cells in significant amounts and the apoptotic cleavage of NuMA was clarified to have an important role in the dissociation of NuMA from the chromatin. The third study concentrated on the interphase function of NuMA showing NuMA depletion to result in cell cycle arrest and the cytoplasmic relocalization of NuMA interaction partner GAS41. We suggest that the relocalization of the transcription factor GAS41 may mediate the cell cycle arrest. Thus, this study has given new aspects in the interactions of NuMA, chromatin and the nuclear matrix.

Passiiviset turvallisuusjärjestelmät kolmannen sukupolven painevesireaktorilaitoksissa

Kandidaatintutkielmassa esitellään kolmannen sukupolven painevesireaktorilaitosten passiivisia turvallisuusjärjestelmiä.

Monte Carlo -reaktorifysiikkalaskennan ja laskennallisen virtausmekaniikan kytkentä kuulakekoreaktorissa

Monte Carlo -reaktorifysiikkakoodit nykyisin käytettävissä olevilla laskentatehoilla tarjoavat mielenkiintoisen tavan reaktorifysiikan ongelmien ratkaisuun. Neljännen sukupolven ydinreaktoreissa käytettävät uudet rakenteet ja materiaalit ovat haasteellisia nykyisiin reaktoreihin suunnitelluille laskentaohjelmille. Tässä työssä Monte Carlo -reaktorifysiikkakoodi ja CFD-koodi yhdistetään kytkettyyn laskentaan kuulakekoreaktorissa, joka on yksi korkealämpötilareaktorityyppi. Työssä käytetty lähestymistapa on uutta maailmankin mittapuussa ajateltuna.