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.

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.

Automatic safety belt usage in 1981 Toyotas. Final report.

National Highway Traffic Safety Administration, Washington, D.C.

Evaluation of passive belts for different size occupants. Final report supplement.

National Highway Traffic Safety Administration, Washington, D.C.

Automatic safety belt systems: changes in owner usage over time in GM Chevettes and VW Rabbits. Final report.

National Highway Traffic Safety Administration, Washington, D.C.

Calspan/Chrysler Research Safety Vehicle - phase III: final technical report. Volume I: executive summary.

National Highway Traffic Safety Administration, Washington, D.C.

Research Safety Vehicle - phase III: final design report. Volume II: appendices.

National Highway Traffic Safety Administration, Washington, D.C.

Development of specifications for passive belt systems. Final report.

National Highway Traffic Safety Administration, Washington, D.C.

Automotive manufacturing processes. Volume II - manufacturing processes for passive restraint systems. Final report.

National Highway Traffic Safety Administration, Office of Research and Development, Washington, D.C.

Automatic safety belt systems owner usage and attitudes in GM Chevettes and VW Rabbits (1980 models). Final report.

National Highway Traffic Safety Administration, Office of Research and Development, Washington, D.C.

Comfort and convenience specifications for safety belts: shoulder belt fit, pressure and pullout forces. Final report.

National Highway Traffic Safety Administration, Washington, D.C.

Evaluation of passive belts for different size occupants. Phase II. Final report.

National Highway Traffic Safety Administration, Washington, D.C.

Evaluation of passive belts for different size occupants. Final report, phase I.

National Highway Traffic Safety Administration, Washington, D.C.

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.

Passiivisten turvallisuusjärjestelmien luontaiset virhetoiminnot ydinvoimalaitoksessa

Uusissa ydinvoimalaitostyypeissä aiotaan käyttää aiempaa enemmän passiivisia turvallisuusjärjestelmiä. Näistä järjestelmistä on vielä vähän käyttökokemusta aktiivisiin turvallisuusjärjestelmiin verrattuna. Työssä tarkastellaan passiivisten turvallisuusjärjestelmien toimintaa sekä etsitään niiden mahdollisia luontaisia vikatilanteita. Luontaisten vikatilanteiden seurauksia järjestelmän suorituskykyyn arvioitiin yksinkertaisilla laskuilla ja mallintamalla RELAP5/MOD3.2.2 beta -termohydrauliikkaohjelmalla. Tarkastelu rajattiin kahden erityyppisen ydinvoimalaitoksen passiivisiin turvallisuusjärjestelmiin. Turvallisuusjärjestelmien suuntaa antavat mitat ja käyttötilanteiden parametrit saatiin laitosvalmistajien laitoskuvauksista. Osoittautui, että vikatilanteissa passiivisissa turvallisuusjärjestelmissä geometrialla on merkittävä vaikutus järjestelmän kapasiteettiin. Tarkasteluissa saatiin myös selville, että laitosmittakaavassa painovoimaisen hätälisävesijärjestelmän turvallisuustoiminto voi toteutua vaikka esiintyisi lyhytaikaisia toimintahäiriöitä, kuten lauhtumista hätälisävesisäiliössä. Sen sijaan lämmönsiirtopiirin virtausreittien tukkeutuminen voi olla fysikaalisesti merkittävä toimintaa haittaava tekijä.