Modelling of a Hydrogen Catalytic Recombiner for Nuclear Power Plant Containment Studies; Case: Siemens FR90/1-150 Recombiner in TONUS OD Code

Ydinvoimalaitokset on suunniteltu ja rakennettu niin, että niillä on kyky selviytyä erilaisista käyttöhäiriöistä ja onnettomuuksista ilman laitoksen vahingoittumista sekä väestön ja ympäristön vaarantumista. On erittäin epätodennäköistä, että ydinvoimalaitosonnettomuus etenee reaktorisydämen vaurioitumiseen asti, minkä seurauksena sydänmateriaalien hapettuminen voi tuottaa vetyä. Jäädytyspiirin rikkoutumisen myötä vety saattaa kulkeutua ydinvoimalaitoksen suojarakennukseen, jossa se voi muodostaa palavan seoksen ilman hapen kanssa ja palaa tai jopa räjähtää. Vetypalosta aiheutuvat lämpötila- ja painekuormitukset vaarantavat suojarakennuksen eheyden ja suojarakennuksen sisällä olevien turvajärjestelmien toimivuuden, joten tehokas ja luotettava vedynhallintajärjestelmä on tarpeellinen. Passiivisia autokatalyyttisiä vetyrekombinaattoreita käytetäänyhä useammissa Euroopan ydinvoimaitoksissa vedynhallintaan. Nämä rekombinaattorit poistavat vetyä katalyyttisellä reaktiolla vedyn reagoidessa katalyytin pinnalla hapen kanssa muodostaen vesihöyryä. Rekombinaattorit ovat täysin passiivisiaeivätkä tarvitse ulkoista energiaa tai operaattoritoimintaa käynnistyäkseen taitoimiakseen. Rekombinaattoreiden käyttäytymisen tutkimisellatähdätään niiden toimivuuden selvittämiseen kaikissa mahdollisissa onnettomuustilanteissa, niiden suunnittelun optimoimiseen sekä niiden optimaalisen lukumäärän ja sijainnin määrittämiseen suojarakennuksessa. Suojarakennuksen mallintamiseen käytetään joko keskiarvoistavia ohjelmia (Lumped parameter (LP) code), moniulotteisia virtausmalliohjelmia (Computational Fluid Dynamics, CFD) tai näiden yhdistelmiä. Rekombinaattoreiden mallintaminen on toteutettu näissä ohjelmissa joko kokeellisella, teoreettisella tai yleisellä (eng. Global Approach) mallilla. Tämä diplomityö sisältää tulokset TONUS OD-ohjelman sisältämän Siemens FR90/1-150 rekombinaattorin mallin vedynkulutuksen tarkistuslaskuista ja TONUS OD-ohjelmalla suoritettujen laskujen tulokset Siemens rekombinaattoreiden vuorovaikutuksista. TONUS on CEA:n (Commissariat à 1'En¬ergie Atomique) kehittämä LP (OD) ja CFD -vetyanalyysiohjelma, jota käytetään vedyn jakautumisen, palamisenja detonaation mallintamiseen. TONUS:sta käytetään myös vedynpoiston mallintamiseen passiivisilla autokatalyyttisillä rekombinaattoreilla. Vedynkulutukseen vaikuttavat tekijät eroteltiin ja tutkittiin yksi kerrallaan. Rekombinaattoreiden vuorovaikutuksia tutkittaessa samaan tilavuuteen sijoitettiin eri kokoisia ja eri lukumäärä rekombinaattoreita. Siemens rekombinaattorimalli TONUS OD-ohjelmassa laskee vedynkulutuksen kuten oletettiin ja tulokset vahvistavat TONUS OD-ohjelman fysikaalisen laskennan luotettavuuden. Mahdollisia paikallisia jakautumia tutkitussa tilavuudessa ei voitu havaita LP-ohjelmalla, koska se käyttäälaskennassa suureiden tilavuuskeskiarvoja. Paikallisten jakautumien tutkintaan tarvitaan CFD -laskentaohjelma.

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.

Nuclear magnetic resonance investigation of structure and dynamics of some electrolyte solutions

Electrolyte solutions are of importance in a wide range of scientific contexts and as such have attracted considerable theoretical and experimental effort over many years. Nuclear Magnetic resonance provides a precise and versatile tool for investigation of electrolyte solutions, both in water and in organic solvents. Many structural and dynamic properties can be obtained through NMR experiments. The solution of aluminum chloride in water was studied. Different concentrations were taken for investigation. Independence of maximum line shift from concentration and acidity was shown. Six-coordinated structure of solvation shell was confirmed by experiments on 'H and 27A1 nuclei. Diffusion coefficients were studied. The solution of nickel chloride in methanol was studied. Lines, corresponding to coordinated and bulk methanol were found. Four-, five- and six-coordinated structures were found in different temperatures. The line for coordinated -OD group of deuterated methanol was observed on 2H spectrum for the first time. Partial deuteration of CH3 group was detected. Inability to observe coordinated -OH group was explained.

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.

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.

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

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.

Requirement management in nuclear power projects

Diplomityön tarkoituksena oli tutkia vaatimusten hallintaa suunnittelu- ja konsultointiyrityksen kannalta Suomen ydinvoimaprojekteissa keskittyen ydinturvallisuus- ja laatuvaatimuksiin. Ydinvoimaprojekteissa toimiminen on edellyttänyt menettelyohjeiden ja laatujärjestelmän uudelleen organisointia yrityksessä ja esiin on noussut haasteita liittyen muun muassa vaatimusten tunnistamiseen ja todentamiseen erityyppisissä ja erilaajuisissa projekteissa. Työ toteutettiin perehtymällä ydinvoimaan liittyvään lainsäädäntöön Suomessa, ohjeisiin ja standardeihin sekä haastattelemalla yrityksen omia asiantuntijoita. Viimeaikaisista sekä meneillään olevista projekteista kerättiin kokemuksia sekä arvioitiin ydinvoima projekteja varten laaditun projektin toteutusohjeen toimivuutta ja käytettävyyttä esimerkkiprojektin avulla. Suurimmiksi haasteiksi tunnistettiin lainsäädännöllisten vaatimusten, kuten ydinvoima- laitosohjeiden (YVL) muuttuminen ja tulkinnanvaraisuus sekä asiakkaiden perehtymät- tömyys Suomen lainsäädäntöön ja vaatimustasoon liittyen ydinturvallisuuteen. Työn tuloksena tunnistettiin hyviä vaatimusten hallintaan liittyviä projektinhallintaa ja ydin- turvallisuutta edistäviä asioita, kuten vaatimusten täsmentäminen jo sopimustasolla sekä niiden täyttymisen seuranta projektin aikana. Erillisen vaatimustietokannan luomista ydinvoimaprojekteja varten tutkittiin, mutta siitä luovuttiin teknisten vaatimusten osalta kannattamattomana, sillä standardien ja vaatimusten määrä kasvoi niin suureksi, että niiden hallitseminen vaatisi enemmän työtä kuin mitä projektien taso yleensä sallisi.

The probabilistic risk analysis of external hazards of an interim storage for spent nuclear fuel

Tässä diplomityössä tehtiin Olkiluodon ydinvoimalaitoksella sijaitsevan käytetyn ydinpolttoaineen allasvarastointiin perustuvan välivaraston todennäköisyysperustainen ulkoisten uhkien riskianalyysi. Todennäköisyysperustainen riskianalyysi (PRA) on yleisesti käytetty riskien tunnistus- ja lähestymistapa ydinvoimalaitoksella. Työn tarkoituksena oli laatia täysin uusi ulkoisten uhkien PRA-analyysi, koska Suomessa ei ole aiemmin tehty vastaavanlaisia tämän tutkimusalueen riskitarkasteluja. Riskitarkastelun motiivina ovat myös maailmalla tapahtuneiden luonnonkatastrofien vuoksi korostunut ulkoisten uhkien rooli käytetyn ydinpolttoaineen välivarastoinnin turvallisuudessa. PRA analyysin rakenne pohjautui tutkimuksen alussa luotuun metodologiaan. Analyysi perustuu mahdollisten ulkoisten uhkien tunnistamiseen pois lukien ihmisen aikaansaamat tahalliset vahingot. Tunnistettujen ulkoisten uhkien esiintymistaajuuksien ja vahingoittamispotentiaalin perusteella ulkoiset uhat joko karsittiin pois tutkimuksessa määriteltyjen karsintakriteerien avulla tai analysoitiin tarkemmin. Tutkimustulosten perusteella voitiin todeta, että tiedot hyvin harvoin tapahtuvista ulkoisista uhista ovat epätäydellisiä. Suurinta osaa näistä hyvin harvoin tapahtuvista ulkoisista uhista ei ole koskaan esiintynyt eikä todennäköisesti koskaan tule esiintymään Olkiluodon vaikutusalueella tai edes Suomessa. Esimerkiksi salaman iskujen ja öljyaltistuksen roolit ja vaikutukset erilaisten komponenttien käytettävyyteen ovat epävarmasti tunnettuja. Tutkimuksen tuloksia voidaan pitää kokonaisuudessaan merkittävinä, koska niiden perusteella voidaan osoittaa ne ulkoiset uhat, joiden vaikutuksia olisi syytä tutkia tarkemmin. Yksityiskohtaisempi tietoisuus hyvin harvoin esiintyvistä ulkoisista uhista tarkentaisi alkutapahtumataajuuksien estimaatteja.

New Type Nuclear Reactors (NETNUC) 2008-2011 Final Report

This report summarizes the work done by a consortium consisting of Lappeenranta University of Technology, Aalto University and VTT Technical Research Centre of Finland in the New Type Nuclear Reactors (NETNUC) project during 2008–2011. The project was part of the Sustainable Energy (SusEn) research programme of the Academy of Finland. A wide range of generation IV nuclear technologies were studied during the project and the research consisted of multiple tasks. This report contains short articles summarizing the results of the individual tasks. In addition, the publications produced and the persons involved in the project are listed in the appendices.