Sviluppo delle tecnologie del circuito a metallo liquido del modulo di blanket HCLL del reattore a fusione ITER e DEMO

Lo scopo della presente tesi di dottorato è di illustrare il lavoro svolto nella progettazione del circuito a metallo liquido del Test Blanket System (TBS) Helium Cooled Lithium Lead (HCLL), uno dei sistemi fondamentali del reattore sperimentale ITER che dovrà dimostrare la fattibilità di produrre industrialmente energia elettrica da processi di fusione nucleare. Il blanket HCLL costituisce una delle sei configurazioni che verranno testate in ITER, sulla base degli esperimenti condotti nei 10 dieci anni di vita del reattore verrà selezionata la configurazione che determinerà la costituzione del primo reattore dimostrativo per la produzione di un surplus di energia elettrica venti volte superiore all’energia consumata, DEMO. Il circuito ausiliario del blanket HCLL è finalizzato, in DEMO all’estrazione del trizio generato mediante il TES; ed in ITER alla dimostrazione della fattibilità di estrarre il trizio generato e di poter gestire il ciclo del trizio. Lo sviluppo dei componenti, svolto in questa tesi, è accentrato su tale dispositivo, il TES. In tale ambito si inseriscono le attività che sono descritte nei capitoli della seguente tesi di dottorato: selezione e progettazione preliminare del sistema di estrazione del trizio dalla lega eutettica Pb15.7Li del circuito a metallo liquido del TBM HCLL; la progettazione, realizzazione e qualifica dei sensori a permeazione per la misura della concentrazione di trizio nella lega eutettica Pb15.7Li; la qualificazione sperimentale all’interno dell’impianto TRIEX (TRItium EXtarction) della tecnologia selezionata per l’estrazione del trizio dalla lega; la progettazione della diagnostica di misura e controllo del circuito ausiliario del TBM HCLL.

Neutronics analyses for fast spectrum nuclear systems and scenario studies for advanced nuclear fuel cycles

The present PhD thesis summarizes the three-years study about the neutronic investigation of a new concept nuclear reactor aiming at the optimization and the sustainable management of nuclear fuel in a possible European scenario. A new generation nuclear reactor for the nuclear reinassance is indeed desired by the actual industrialized world, both for the solution of the energetic question arising from the continuously growing energy demand together with the corresponding reduction of oil availability, and the environment question for a sustainable energy source free from Long Lived Radioisotopes and therefore geological repositories. Among the Generation IV candidate typologies, the Lead Fast Reactor concept has been pursued, being the one top rated in sustainability. The European Lead-cooled SYstem (ELSY) has been at first investigated. The neutronic analysis of the ELSY core has been performed via deterministic analysis by means of the ERANOS code, in order to retrieve a stable configuration for the overall design of the reactor. Further analyses have been carried out by means of the Monte Carlo general purpose transport code MCNP, in order to check the former one and to define an exact model of the system. An innovative system of absorbers has been conceptualized and designed for both the reactivity compensation and regulation of the core due to cycle swing, as well as for safety in order to guarantee the cold shutdown of the system in case of accident. Aiming at the sustainability of nuclear energy, the steady-state nuclear equilibrium has been investigated and generalized into the definition of the ``extended'' equilibrium state. According to this, the Adiabatic Reactor Theory has been developed, together with a New Paradigm for Nuclear Power: in order to design a reactor that does not exchange with the environment anything valuable (thus the term ``adiabatic''), in the sense of both Plutonium and Minor Actinides, it is required indeed to revert the logical design scheme of nuclear cores, starting from the definition of the equilibrium composition of the fuel and submitting to the latter the whole core design. The New Paradigm has been applied then to the core design of an Adiabatic Lead Fast Reactor complying with the ELSY overall system layout. A complete core characterization has been done in order to asses criticality and power flattening; a preliminary evaluation of the main safety parameters has been also done to verify the viability of the system. Burn up calculations have been then performed in order to investigate the operating cycle for the Adiabatic Lead Fast Reactor; the fuel performances have been therefore extracted and inserted in a more general analysis for an European scenario. The present nuclear reactors fleet has been modeled and its evolution simulated by means of the COSI code in order to investigate the materials fluxes to be managed in the European region. Different plausible scenarios have been identified to forecast the evolution of the European nuclear energy production, including the one involving the introduction of Adiabatic Lead Fast Reactors, and compared to better analyze the advantages introduced by the adoption of new concept reactors. At last, since both ELSY and the ALFR represent new concept systems based upon innovative solutions, the neutronic design of a demonstrator reactor has been carried out: such a system is intended to prove the viability of technology to be implemented in the First-of-a-Kind industrial power plant, with the aim at attesting the general strategy to use, to the largest extent. It was chosen then to base the DEMO design upon a compromise between demonstration of developed technology and testing of emerging technology in order to significantly subserve the purpose of reducing uncertainties about construction and licensing, both validating ELSY/ALFR main features and performances, and to qualify numerical codes and tools.

Development, mechanical characterization and qualification in liquid lead of stainless steels for structural applications in the future fission and fusion nuclear reactors

The research activity carried out in the Brasimone Research Center of ENEA concerns the development and mechanical characterization of steels conceived as structural materials for future fission reactors (Heavy Liquid Metal IV Generation reactors: MYRRHA and ALFRED) and for the future fusion reactor DEMO. Within this framework, two parallel lines of research have been carried out: (i) characterization in liquid lead of steels and weldings for the components of the IV Generation fission reactors (GIV) by means of creep and SSRT (Slow Strain Rate Tensile) tests; (ii) development and screening on mechanical properties of RAFM (Reduced Activation Ferritic Martensitic) steels to be employed as structural materials of the future DEMO fusion reactor. The doctoral work represents therefore a comprehensive report of the research carried out on nuclear materials both from the point of view of the qualification of existing (commercial) materials for their application in the typical environmental conditions of 4th generation fission reactors operating with lead as coolant, and from the point of view of the metallurgical study (with annexed microstructural and mechanical characterization of the selected compositions / Thermo Mechanical Treatment (TMT) options) of new compositional variants to be proposed for the “Breeding Blanket” of the future DEMO Fusion Reactor.