Suchergebnisse

This PhD thesis is evaluating the safety level of the graphite-moderated boiling water cooled nuclear power reactors (RBMK reactors) by the use of best estimate three dimensional neutron kinetics coupled thermal-hydraulics codes. The availability of such sophisticated tools has allowed detailed and realistic analyses of these kind of reactors, also known as "Chernobyl-type" reactors. Chernobyl is the name of a RBMK reactor where, in 1986, a severe accident occurred, leading to the destruction of the plant and to a major release of radioactivity into the environment. Parts of the activities of this PhD thesis were developed in the framework of the European Union funded TACIS Project R2.03/97 "Software development for the RBMK and WWER reactors". This project was awarded to the "Gruppo di Ricerca Nucleare San Piero a Grado" of the University of Pisa and managed by it in collaboration with the RBMK designers (NIKIET, Kurchatov Insititute) and the licensee (RosEnergoAtom, now EnergoAtom Concern OJSC). The research activities dealt with the development and the validation of a sophisticated thermal-hydraulic nodalization of the Smolensk-3 Nuclear Power Plant. This thermal-hydraulic model was then coupled with a three dimensional neutron kinetics model of the core. The code used was RELAP5-3D system code. Suitable RBMK cross sections libraries were developed in collaboration with the Pennsylvania State University, using the deterministic lattice physics code HELIOS. After the validation of the developed models, the most relevant transients for the plant safety at full power were calculated, e.g. the group distribution header rupture, the break of the control and protection system cooling circuit. A special emphasis was put in the simulation of the single fuel channel transient, using also the Monte Carlo code MCNP5. The last part of the PhD activities concerned the analysis of a low power transient. In particular, the Chernobyl extreme scenario was reconstructed. Xenon fuel cell cross sections were calculated using the deterministic transport code DRAGON. Finally, all the analyses performed in the framework of this PhD confirmed the upgraded level of nuclear safety of the RBMK reactors, obtained also as a consequence of the relevant hardware modifications implemented in the aftermath of the Chernobyl accident.

Problems about future energy availability, climate changes, and air quality seem to play an important role in energy production. While current reactor generations provide a guaranteed and economical energy production, new nuclear power plant generation would increase the ways and purposes in which nuclear energy can be used. To explore these new technological applications, several governments, industries, and research communities decided to contribute to the next reactor generation, called "Generation IV." Among the six Gen-IV reactor designs, the Gas Cooled Fast Reactor (GCFR) uses a direct-cycle helium turbine for electricity generation and for a -free thermochemical production of hydrogen. Additionally, the use of a fast spectrum allows actinides transmutation, minimizing the production of long-lived radioactive waste in an integrated fuel cycle. This paper presents an analysis of GCFR fuel cycle optimization and of a thermal-hydraulic of a GCFR-prototype under steady-state and transient conditions. The fuel cycle optimization was performed to assess the capability of the GCFR to transmute MAs, while the thermal-hydraulic analysis was performed to investigate the reactor and the safety systems behavior during a LOFA. Preliminary results show that limited quantities of MA are not affecting significantly the thermal-fluid-dynamics behavior of a GCFR core.