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.

VVER-440 Thermal Hydraulics as a Computer Code Validation Challenge

This thesis concentrates on the validation of a generic thermal hydraulic computer code TRACE under the challenges of the VVER-440 reactor type. The code capability to model the VVER-440 geometry and thermal hydraulic phenomena specific to this reactor design has been examined and demonstrated acceptable. The main challenge in VVER-440 thermal hydraulics appeared in the modelling of the horizontal steam generator. The major challenge here is not in the code physics or numerics but in the formulation of a representative nodalization structure. Another VVER-440 specialty, the hot leg loop seals, challenges the system codes functionally in general, but proved readily representable. Computer code models have to be validated against experiments to achieve confidence in code models. When new computer code is to be used for nuclear power plant safety analysis, it must first be validated against a large variety of different experiments. The validation process has to cover both the code itself and the code input. Uncertainties of different nature are identified in the different phases of the validation procedure and can even be quantified. This thesis presents a novel approach to the input model validation and uncertainty evaluation in the different stages of the computer code validation procedure. This thesis also demonstrates that in the safety analysis, there are inevitably significant uncertainties that are not statistically quantifiable; they need to be and can be addressed by other, less simplistic means, ultimately relying on the competence of the analysts and the capability of the community to support the experimental verification of analytical assumptions. This method completes essentially the commonly used uncertainty assessment methods, which are usually conducted using only statistical methods.

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.

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.

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.

Ydinvoimalaitoksen primääripiiriä kuvaavan PACTEL-koelaitteiston mallinnus ja kelpoistus APROS 5 simulointiympäristössä.

Ydinvoimalaitosten turvallisuusanalyysit tehdään nykyisin pääasiassa tietokoneohjelmistoilla. Turvallisuusanalyyseissä käytetyt ohjelmistot ja niillä tehdyt mallit pitää kelpoistaa, jotta mallilla saatuja tuloksia voidaan pitää luotettavina. PACTEL-koelaitteistolla tehdään turvallisuustutkimusta, joka palvelee erityisesti Loviisan VVER-440 -tyyppisiä voimalaitoksia. APROS-koodi kehitettiin Loviisan voimalaitoksen turvallisuusanalyysejä varten. Jotta APROS-koodi voitaisiin kelpoistaa rakennettiin PACTEL-koelaitteisto kokeellista termohydrauliikkatutkimusta varten. Koelaitteiston tuloksia käytettiin APROS ohjelmiston termohydraulisten mallien kehittämiseen. Vuonna 1999 aloitetun kansallisen FINNUS-projektin osatavoite on kehittää turvallisuustutkimuksissa käytettyjä tietokoneohjelmia, kuten APROSia. APROS on kehittynyt vuosien varrella niin laskenta-algoritmien kuin fysikaalisten mallienkin osalta. APROSiin oli kehitetty myös uusi käyttöliittymä GRADES, joka toimii Windows NT-ympäristössä. Diplomityön tavoitteena oli tehdä uudella GRADES-käyttöliittymällä uusi ja entistä tarkempi simulaatiomalli PACTEL-koelaitteistosta. Uusi simulaatiomalli kelpoistettiin kahden vanhan PACTEL-kokeen avulla, LOF-10 ja SBL-22. Laskentatuloksista voidaan päätellä laskeeko APROS oikein ja voidaanko APROSilla tehtyjä turvallisuusanalyysejä pitää luotettavina. Valmis kelpoistettu simulaatiomalli tuli VTT Energian kokeellisen lämpö- ja virtaustekniikan laboratorion käyttöön. Simulaatiomallilla voidaan laskea ja simuloida sekä vanhoja että uusia PACTEL-kokeita ja käyttää mallia tulevien PACTEL-kokeiden suunnitteluun.

The Effect of Load Follow Operation on Fuel Behaviour in Loviisa NPP

Työn tarkoituksena oli analysoida polttoainesauvojen käyttäytymistä Loviisan ydinvoimalaitoksen tehonsäätöajossa. Sähkömarkkinoiden vapautuminen Pohjoismaissa sekä tämän seurauksena vaihteleva sähkön markkinahinta ovat ajaneet sähkötuottajat tilanteeseen, jossa tuotanto aiempaa enemmän mukautuu markkinatilanteeseen. Näin ollen myös Loviisan ydinvoimalaitoksen osallistuminen sähkön tuotannon säätelyyn saattaa tulevaisuudessa olla ajankohtaista. Ennen kuin reaktorin tehonsäätöajoa voidaan alkaa toteuttaa, tulee varmistua siitä, että polttoainesauvassa tehonsäätöjen seurauksena tapahtuvat muutokset eivät aiheuta epäsuotuisia käyttäytymisilmiöitä. Työssä tarkastellaan kahden Loviisan ydinvoimalaitoksen polttoainetoimittajan, British Nuclear Fuels plc:n ja venäläisen TVEL:n ensinippujen polttoainesauvan käyttäytymistä tehonsäätötapauksissa. Työssä tarkastellut tehonsäätötapaukset on pyritty valitsemaan niin, että ne kuvaisivat tulevaisuudessa mahdollisesti toteutettavia tehonsäätöjä. Laskentatapauksien sauvatehohistoriat on generoitu HEXBU-3D sydänsimulaattoriohjelmalla lasketun nelivuotisen perustehohistorian pohjalta lisäämällä säätösauvan aiheuttama reaktoritehon muutos, säätösauvan viereisen polttoainenipun aksiaalitehon muutos sekä säätösauvan rakenteen aiheuttama paikallinen tehopiikki säätösauvan vieressä. Työssä tarkastellaan tehonsäätöjen toteuttamista eri tehotasoille ja vaihtelevilla määrillä tehonsäätösyklejä. Työssä käsitellyt laskentatapaukset on jaoteltu reaktorin ajotavan mukaan seuraavasti: peruskuorma-ajo, viikonloppusäätö ja päiväsäätö. Laskenta suoritettiin ydinpolttoaineen käyttäytymistä kuvaavaa ENIGMA-B 7.3.0 ohjelmaa apuna käyttäen. Laskelmien tulokset osoittavat, että molempien polttoainetoimittajien ensinippujen sauvat kestävät reaktorin tehonsäätöajoa rajoituksetta tarkastelluissa laskentatapauksissa. ENIGMA-ohjelman sisältämät mallit, jotka ennustavat polttoainesauvan suojakuoren vaurioitumistodennäköisyyden jännityskorroosion tai väsymismurtuman kautta, eivät näytä mitään merkkejä vaurioitumisesta. BNFL:n polttoainesauva saavuttaa kuitenkin suurempia väsymismurtumatodennäköisyyden arvoja. Tämä johtuu siitä, että polttoainepelletin ja suojakuoren välinen mekaaninen vuorovaikutus syntyy BNFL:n sauvassa aikaisemmin, joka taas johtaa suurempaan määrään sauvaa rasittavia muodonmuutoksia tehonnostotilanteissa. TVEL:n Zr1%Nb -materiaalista valmistetun suojakuoren käyttäytymistä ei voida kuitenkaan suoraan näiden laskujen perusteella arvioida, sillä ENIGMA-ohjelman mallit perustuvat Zircaloy-suojakuorimateriaaleilla suoritettuihin kokeisiin.

Prosessihyötysuhteen parantamiskohteiden kartoitus painevesireaktorityyppisessä ydinvoimalaitoksessa

Työn tavoitteena on kartoittaa painevesireaktorityyppisen ydinvoimalaitoksen prosessihyötysuhteen parantamiskohteita. Aluksi kirjallisuudesta etsitään hyötysuhteen parantamiskeinoja ideaalisessa höyryvoimalaitosprosessissa. Näistä valitaan sopivimmat tarkastelun kohteeksi todellisessa voimalaitoksessa: syöttöveden esilämmityksen tehostaminen väliottohöyryvirtausta kasvattamalla ja syöttöveden esilämmittimen lämmönsiirtopintaa lisäämällä. Tarkastelussa pyritään löytämään paras mahdollinen hyötysuhde väliottohöyrylinjojen putkikokoa sekä esilämmittimien putkien lukumäärää muuttamalla. Diskreetin optimoinnin iteraatioaskel määritetään hyötysuhteen osittaisderivaattojen avulla. Tehtäviä muutoksia simuloidaan APROS-simulointiohjelmalla, jossa käytetään Loviisan voimalaitoksesta tehtyä mallia VVER-440. Työssä havaittiin, että pelkkiä väliottohöyrylinjojen putkikokoja – ja massavirtaa – kasvattamalla Loviisan voimalaitoksen hyötysuhdetta voidaan parantaa parhaimmillaan 32,75%:sta 32,85%:iin. Syöttöveden esilämmittimien lämmönsiirtopintaa lisäämällä saadaan suurempi parannus hyötysuhteeseen: 32,75%:sta 32,99%:iin. Näissä tapauksissa muutettiin kaikkia väliottohöyrylinjoja tai syöttöveden esilämmittimien lämpöpintoja. Työssä tarkasteltiin myös joitakin pienempiä muutoskohteita, joista paras hyötysuhteen kasvu saatiin korkeapaine-esilämmittimien lämmönsiirtopintaa kasvattamalla sekä toisen väliottohöyrylinjan (RD12) ja sitä vastaavan syöttöveden esilämmittimen muutosten yhteisvaikutuksena.

USER'S GUIDE FOR ADVANCED PROCESS SIMULATION SOFTWARE APROS 5

Tässä diplomityössä tehtiin käyttäjän opas kehittyneelle prosessisimulointiohjelmistolle APROS 5. Opas on osa VTT Energialle tehtävää APROS 5 käyttäjän koulutuspakettia, joka julkaistaan myöhemmin CD-ROM -muotoisena. Prosessisimulointiohjelmistoa AAPROS 5 voidaan käyttää termohydraulisten prosessien, automaatiopiirien ja sähköjärjestelmien mallinnuksessa. Ohjelma sisältää myös neutroniikkamallin ydinreaktorin käyttäytymisen mallintamiseksi. APROS:in aikaisemmilla UNIX-ympäristössä toimivilla versioilla on toteutettu useita ydinvoimalaitosten turvallisuustutkimukseen liittyviä analyysejä ja sekä ydinvoimalaitosten että konventionaalisten voimalaitosten koulutussimulaattoreita. APROS 5 toimii Windows NT -ympäristössä ja on oleellisesti erilainen käyttää kuin aikaisemmat versiot. Tämän myötä syntyi tarve uudelle käyttäjän oppaalle. Käyttäjän oppaassa esitetään APROS 5:n tärkeimmät toiminnot, mallinnuksen periaatteet ja termohydraulisten ja neutroniikan ratkaisumallit. Lisäksi oppaassa esitetään esimerkki, jossa mallinnetaan yksinkertaistettu VVER-440 -tyyppisen ydinvoimalaitoksen primääripiiri. Yksityiskohtaisempaa tietoa ohjelmistosta on saatavilla APROS 5 -dokumentaatiosta.

Ydinreaktorin painesäiliön ulkoisen jäähdytyksen lämmönsiirtokokeita

The purpose of this work was to design and carry out thermal-hydraulic experiments dealing with overcooling transients of a VVER-440-type nuclear reactor pressure vessel. Sudden overcooling accident could have negative effect on the mechanical strength of the pressure vessel. If part of the pressure vessel is compromised, the intense pressure inside a pressurized water reactor could cause the wall to fracture. Information on the heat transfer along the outside of the pressure vessel wall is necessary for stress analysis. Basic knowledge of the overcooling accident and heat transfer types on the outside of the pressure vessel is presented as background information. Test facility was designed and built based to study and measure heat transfer during specific overcooling scenarios. Two test series were conducted with the first one concentrating on the very beginning of the transient and the second one concentrating on steady state heat transfer. Heat transfer coefficients are calculated from the test data using an inverse method, which yields better results in fast transients than direct calculation from the measurement results. The results show that heat transfer rate varies considerably during the transient, being very high in the beginning and dropping to steady state in a few minutes. The test results show that appropriate correlations can be used in future analysis.

Implementation of population balances in simulation of gas-liquid reactors using a CFD-PBM coupled approach

In bubbly flow simulations, bubble size distribution is an important factor in determination of hydrodynamics. Beside hydrodynamics, it is crucial in the prediction of interfacial area available for mass transfer and in the prediction of reaction rate in gas-liquid reactors such as bubble columns. Solution of population balance equations is a method which can help to model the size distribution by considering continuous bubble coalescence and breakage. Therefore, in Computational Fluid Dynamic simulations it is necessary to couple CFD and Population Balance Model (CFD-PBM) to get reliable distribution. In the current work a CFD-PBM coupled model is implemented as FORTRAN subroutines in ANSYS CFX 10 and it has been tested for bubbly flow. This model uses the idea of Multi Phase Multi Size Group approach which was previously presented by Sha et al. (2006) [18]. The current CFD-PBM coupled method considers inhomogeneous flow field for different bubble size groups in the Eulerian multi-dispersed phase systems. Considering different velocity field for bubbles can give the advantageof more accurate solution of hydrodynamics. It is also an improved method for prediction of bubble size distribution in multiphase flow compared to available commercial packages.

Flow phenomena, heat and mass transfer in microchannel reactors

The studies of flow phenomena, heat and mass transfer in microchannel reactors are beneficial to estimate and evaluate the ability of microchannel reactors to be operated for a given process reaction such as Fischer-Tropsch synthesis. The flow phenomena, for example, the flow regimes and flow patterns in microchannel reactors for both single phase and multiphase flow are affected by the configuration of the flow channel. The reviews of the previous works about the analysis of related parameters that affect the flow phenomena are shown in this report. In order to predict the phenomena of Fischer-Tropsch synthesis in microchannel reactors, the 3-dimensional computational fluid dynamic simulation with commercial software package FLUENT was done to study the flow phenomena and heat transfer for gas phase Fischer-Tropsch products flow in rectangular microchannel with hydraulic diameter 500 ¿m and length 15 cm. Numerical solution with slip boundary condition was used in the simulation and the flowphenomena and heat transfer were determined.

Selective catalytic reduction of nitrogen oxides with ammonia in forced unsteady state reactors - Case based reasoning and mathematical model simulation reasoning

The application of forced unsteady-state reactors in case of selective catalytic reduction of nitrogen oxides (NOx) with ammonia (NH3) is sustained by the fact that favorable temperature and composition distributions which cannot be achieved in any steady-state regime can be obtained by means of unsteady-state operations. In a normal way of operation the low exothermicity of the selective catalytic reduction (SCR) reaction (usually carried out in the range of 280-350°C) is not enough to maintain by itself the chemical reaction. A normal mode of operation usually requires supply of supplementary heat increasing in this way the overall process operation cost. Through forced unsteady-state operation, the main advantage that can be obtained when exothermic reactions take place is the possibility of trapping, beside the ammonia, the moving heat wave inside the catalytic bed. The unsteady state-operation enables the exploitation of the thermal storage capacity of the catalyticbed. The catalytic bed acts as a regenerative heat exchanger allowing auto-thermal behaviour when the adiabatic temperature rise is low. Finding the optimum reactor configuration, employing the most suitable operation model and identifying the reactor behavior are highly important steps in order to configure a proper device for industrial applications. The Reverse Flow Reactor (RFR) - a forced unsteady state reactor - corresponds to the above mentioned characteristics and may be employed as an efficient device for the treatment of dilute pollutant mixtures. As a main disadvantage, beside its advantages, the RFR presents the 'wash out' phenomena. This phenomenon represents emissions of unconverted reactants at every switch of the flow direction. As a consequence our attention was focused on finding an alternative reactor configuration for RFR which is not affected by the incontrollable emissions of unconverted reactants. In this respect the Reactor Network (RN) was investigated. Its configuration consists of several reactors connected in a closed sequence, simulating a moving bed by changing the reactants feeding position. In the RN the flow direction is maintained in the same way ensuring uniformcatalyst exploitation and in the same time the 'wash out' phenomena is annulated. The simulated moving bed (SMB) can operate in transient mode giving practically constant exit concentration and high conversion levels. The main advantage of the reactor network operation is emphasizedby the possibility to obtain auto-thermal behavior with nearly uniformcatalyst utilization. However, the reactor network presents only a small range of switching times which allow to reach and to maintain an ignited state. Even so a proper study of the complex behavior of the RN may give the necessary information to overcome all the difficulties that can appear in the RN operation. The unsteady-state reactors complexity arises from the fact that these reactor types are characterized by short contact times and complex interaction between heat and mass transportphenomena. Such complex interactions can give rise to a remarkable complex dynamic behavior characterized by a set of spatial-temporal patterns, chaotic changes in concentration and traveling waves of heat or chemical reactivity. The main efforts of the current research studies concern the improvement of contact modalities between reactants, the possibility of thermal wave storage inside the reactor and the improvement of the kinetic activity of the catalyst used. Paying attention to the above mentioned aspects is important when higher activity even at low feeding temperatures and low emissions of unconverted reactants are the main operation concerns. Also, the prediction of the reactor pseudo or steady-state performance (regarding the conversion, selectivity and thermal behavior) and the dynamicreactor response during exploitation are important aspects in finding the optimal control strategy for the forced unsteady state catalytic tubular reactors. The design of an adapted reactor requires knowledge about the influence of its operating conditions on the overall process performance and a precise evaluation of the operating parameters rage for which a sustained dynamic behavior is obtained. An apriori estimation of the system parameters result in diminution of the computational efforts. Usually the convergence of unsteady state reactor systems requires integration over hundreds of cycles depending on the initial guess of the parameter values. The investigation of various operation models and thermal transfer strategies give reliable means to obtain recuperative and regenerative devices which are capable to maintain an auto-thermal behavior in case of low exothermic reactions. In the present research work a gradual analysis of the SCR of NOx with ammonia process in forced unsteady-state reactors was realized. The investigation covers the presentationof the general problematic related to the effect of noxious emissions in the environment, the analysis of the suitable catalysts types for the process, the mathematical analysis approach for modeling and finding the system solutions and the experimental investigation of the device found to be more suitable for the present process. In order to gain information about the forced unsteady state reactor design, operation, important system parameters and their values, mathematical description, mathematicalmethod for solving systems of partial differential equations and other specific aspects, in a fast and easy way, and a case based reasoning (CBR) approach has been used. This approach, using the experience of past similarproblems and their adapted solutions, may provide a method for gaining informations and solutions for new problems related to the forced unsteady state reactors technology. As a consequence a CBR system was implemented and a corresponding tool was developed. Further on, grooving up the hypothesis of isothermal operation, the investigation by means of numerical simulation of the feasibility of the SCR of NOx with ammonia in the RFRand in the RN with variable feeding position was realized. The hypothesis of non-isothermal operation was taken into account because in our opinion ifa commercial catalyst is considered, is not possible to modify the chemical activity and its adsorptive capacity to improve the operation butis possible to change the operation regime. In order to identify the most suitable device for the unsteady state reduction of NOx with ammonia, considering the perspective of recuperative and regenerative devices, a comparative analysis of the above mentioned two devices performance was realized. The assumption of isothermal conditions in the beginningof the forced unsteadystate investigation allowed the simplification of the analysis enabling to focus on the impact of the conditions and mode of operation on the dynamic features caused by the trapping of one reactant in the reactor, without considering the impact of thermal effect on overall reactor performance. The non-isothermal system approach has been investigated in order to point out the important influence of the thermal effect on overall reactor performance, studying the possibility of RFR and RN utilization as recuperative and regenerative devices and the possibility of achieving a sustained auto-thermal behavior in case of lowexothermic reaction of SCR of NOx with ammonia and low temperature gasfeeding. Beside the influence of the thermal effect, the influence of the principal operating parameters, as switching time, inlet flow rate and initial catalyst temperature have been stressed. This analysis is important not only because it allows a comparison between the two devices and optimisation of the operation, but also the switching time is the main operating parameter. An appropriate choice of this parameter enables the fulfilment of the process constraints. The level of the conversions achieved, the more uniform temperature profiles, the uniformity ofcatalyst exploitation and the much simpler mode of operation imposed the RN as a much more suitable device for SCR of NOx with ammonia, in usual operation and also in the perspective of control strategy implementation. Theoretical simplified models have also been proposed in order to describe the forced unsteady state reactors performance and to estimate their internal temperature and concentration profiles. The general idea was to extend the study of catalytic reactor dynamics taking into account the perspectives that haven't been analyzed yet. The experimental investigation ofRN revealed a good agreement between the data obtained by model simulation and the ones obtained experimentally.

VVER-640 -reaktorityypin passiiviset turvallisuusominaisuudet

Turvallisuussuunnittelu muodostaa merkittävän osan ydinvoimalaitoksen suunnit-telutyöstä. Uusissa laitoskonsepteissa turvallisuutta on pyritty parantamaan lisää-mällä perinteisten aktiivisten hätäjärjestelmien rinnalle passiivisia eli toiminnal-taan puhtaasti luonnonlakeihin perustuvia hätäjärjestelmiä. Sähköteholtaan 640 MW oleva VVER-640 -laitostyyppi edustaa tässä suhteessa viimeisintä kehi-tysaskelta venäläisten VVER kevytvesireaktorien sarjassa. Suunnittelun lähtökoh-tana on ollut turvallisuuden parantaminen verrattuna aikaisempiin VVER-malleihin. Tähän on pyritty hätäjärjestelmien passiivisella toteutuksella. Passiivis-ten järjestelmien mitoitusperusteena on ollut laitoksen selviäminen itsenäisesti 24 tunnin ajan mahdollisissa onnettomuustilanteissa ilman suojarakennuksen tiiviy-den menetystä. Relap5-ohjelmalla tehtyjen simulointien perusteella laitoksen pas-siiviset järjestelmät näyttäisivät pystyvän huolehtimaan laitoksen turvallisuudesta sekä jäähdytteen- että sähkönmenetysonnettomuuksissa ilman aktiivisten järjes-telmien apua vaaditut 24 tuntia.