Analysis of PWR-PACTEL Small Break LOCA experiment using TRACE

Currently, the power generation is one of the most significant life aspects for the whole man-kind. Barely one can imagine our life without electricity and thermal energy. Thus, different technologies for producing those types of energy need to be used. Each of those technologies will always have their own advantages and disadvantages. Nevertheless, every technology must satisfy such requirements as efficiency, ecology safety and reliability. In the matter of the power generation with nuclear energy utilization these requirements needs to be highly main-tained, especially since accidents on nuclear power plants may cause very long term deadly consequences. In order to prevent possible disasters related to the accident on a nuclear power plant strong and powerful algorithms were invented in last decades. Such algorithms are able to manage calculations of different physical processes and phenomena of real facilities. How-ever, the results acquired by the computing must be verified with experimental data.

The role of nuclear and other conventional power plants in the flexible energy system

This thesis reviews the role of nuclear and conventional power plants in the future energy system. The review is done by utilizing freely accesible publications in addition to generating load duration and ramping curves for Nordic energy system. As the aim of the future energy system is to reduce GHG-emissions and avoid further global warming, the need for flexible power generation increases with the increased share of intermittent renewables. The goal of this thesis is to offer extensive understanding of possibilities and restrictions that nuclear power and conventional power plants have regarding flexible and sustainable generation. As a conclusion, nuclear power is the only technology that is able to provide large scale GHG-free power output variations with good ramping values. Most of the currently operating plants are able to take part in load following as the requirement to do so is already required to be included in the plant design. Load duration and ramping curves produced prove that nuclear power is able to cover most of the annual generation variation and ramping needs in the Nordic energy system. From the conventional power generation methods, only biomass combustion can be considered GHG-free because biomass is considered carbon neutral. CFB combusted biomass has good load follow capabilities in good ramping and turndown ratios. All the other conventional power generation technologies generate GHG-emissions and therefore the use of these technologies should be reduced.

Development of laboratory exercises for basic nuclear thermal hydraulics measurements

The thesis focuses on light water reactors (pressurized water reactors, boiling water reactors) and measurement techniques for basic thermal hydraulics parameters that are used in a nuclear power plant. The goal of this work is a development of laboratory exercises for basic nuclear thermal hydraulics measurements.

Vakavat reaktorionnettomuudet ja niiden mallintaminen Olkiluoto 1 ja Olkiluoto 2 laitosyksiköille MELCOR-ohjelman versiolla 2.1

Vakaviin reaktorionnettomuuksiin liittyviä ilmiöitä on tutkittu jo 1980-luvulta lähtien ja tutkitaan edelleen. Ilmiöt liittyvät reaktorisydämen ja muiden paineastian sisäisten materi-aalien sulamiseen sekä reagointiin veden ja höyryn kanssa. Ilmiöt on myös tärkeää tuntea ja niiden esiintymistä mallintaa käytössä olevilla laitoksilla, jotta voidaan varmistua turval-lisuusjärjestelmien riittävyydestä. Olkiluoto 1 ja 2 laitosten käyttölupa uusitaan vuoteen 2018 mennessä. Lupaprosessiin liit-tyy analyysejä, joissa mallinnetaan laitosten toimintaa vakavassa reaktorionnettomuudessa. Näiden analyysien tekoon Teollisuuden Voima Oyj on käyttänyt ohjelmaa nimeltä MEL-COR jo vuodesta 1994 lähtien. Käytössä on ollut useita eri ohjelmaversioita ja viimeisin niistä on 1.8.6, joka riittää vielä tulevan käyttöluvan uusintaprojektiin liittyvien analyysien tekoon. MELCOR:n vanhaa 1.8.6 ohjelmaversioita ei kuitenkaan enää päivitetä, joten siirtyminen uudempaan 2.1 versioon on tulevaisuudessa välttämätöntä. Uusimman versiopäivityksen yhteydessä on kuitenkin muuttunut koko ohjelman lähdekoodi ja vanhojen laitosmallien käyttö uudessa ohjelmaversiossa vaatii tiedostojen konvertoinnin. Tässä työssä esitellään MELCOR-version 2.1 ominaisuuksia ja selvitetään, mitä 1.8.6 versioon luotujen laitosmal-lien käyttöönotto versiossa 2.1 vaatii. Vaatimusten määrittelemiseksi laitosmalleilla tehdään ajoja molemmilla ohjelmaversioilla ja erilaisilla onnettomuuden alkutapahtuman määrittelyillä. Tulosten perusteella arvioidaan ohjelmaversioiden eroja ja pohditaan mitä puutteita laitosmalleihin konversion jälkeen jää. Näiden perusteella arvioidaan mitä jatkotoimenpiteitä konversio vaatii.

Failure modes of passive decay heat removing safety systems of modern nuclear power plants

The purpose of this master’s thesis is to gain an understanding of passive safety systems’ role in modern nuclear reactors projects and to research the failure modes of passive decay heat removal safety systems which use phenomenon of natural circulation. Another purpose is to identify the main physical principles and phenomena which are used to establish passive safety tools in nuclear power plants. The work describes passive decay heat removal systems used in AES-2006 project and focuses on the behavior of SPOT PG system. The descriptions of the main large-scale research facilities of the passive safety systems of the AES-2006 power plant are also included. The work contains the calculations of the SPOT PG system, which was modeled with thermal-hydraulic system code TRACE. The dimensions of the calculation model are set according to the dimensions of the real SPOT PG system. In these calculations three parameters are investigated as a function of decay heat power: the pressure of the system, the natural circulation mass flow rate around the closed loop, and the level of liquid in the downcomer. The purpose of the calculations is to test the ability of the SPOT PG system to remove the decay heat from the primary side of the nuclear reactor in case of failure of one, two, or three loops out of four. The calculations show that three loops of the SPOT PG system have adequate capacity to provide the necessary level of safety. In conclusion, the work supports the view that passive systems could be widely spread in modern nuclear projects.