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.

Elliptisten käyrien kryptosysteemit: Weierstrassin, Hessen ja Edwardsin käyrien vertailua

Elliptisten käyrien salakirjoitusjärjestelmät (ECC) ovat julkisen avaimen salakirjoitusjärjestelmiä, jotka perustuvat elliptisen käyrän pisteiden muodostamaan ryhmään. Ne ovat tällä hetkellä suuren mielenkiinnon kohteina, sillä perinteisesti käytössä olevan RSA-salausalgoritmin avainten pituudet ovat alkaneet kasvaa epäkäytännöllisen pitkiksi. Elliptisten käyrien järjestelmillä on mahdollista päästä 160 bitin avaimenpituudella vastaavan tasoiseen turvallisuuteen kuin 1024-bittisellä RSA-salausavaimella. Elliptisiin käyriin perustuvilla järjestelmillä on kuitenkin ollut omat ongelmansa yhtenäisyyden kanssa. Tämä on aiheuttanut haasteita salausjärjestelmän rakentamisessa ja saattanut vuotaa sivukanavatietoa. Määrittelemällä käytettävä elliptinen käyrä sopivasti näitä ongelmia on kuitenkin mahdollista välttää. Lisähyötynä voidaan saavuttaa nopeampi laskenta tinkimättä turvallisuudesta. Tässä työssä esitellään elliptisistä käyristä Weierstrassin normaalimuoto, Hessen käyrä, Edwardsin käyrä sekä binääriset Weierstrassin ja Edwardsin käyrät. Kaikille käyrille määritellään pisteiden yhteenlasku sekä esitetään käyräkohtaista tietoa niiden tehokkuudesta laskennassa.