4H silicon carbide particle detectors: study of the defects induced by high energy neutron irradiation

During the last decade advances in the field of sensor design and improved base materials have pushed the radiation hardness of the current silicon detector technology to impressive performance. It should allow operation of the tracking systems of the Large Hadron Collider (LHC) experiments at nominal luminosity (1034 cm-2s-1) for about 10 years. The current silicon detectors are unable to cope with such an environment. Silicon carbide (SiC), which has recently been recognized as potentially radiation hard, is now studied. In this work it was analyzed the effect of high energy neutron irradiation on 4H-SiC particle detectors. Schottky and junction particle detectors were irradiated with 1 MeV neutrons up to fluence of 1016 cm-2. It is well known that the degradation of the detectors with irradiation, independently of the structure used for their realization, is caused by lattice defects, like creation of point-like defect, dopant deactivation and dead layer formation and that a crucial aspect for the understanding of the defect kinetics at a microscopic level is the correct identification of the crystal defects in terms of their electrical activity. In order to clarify the defect kinetic it were carried out a thermal transient spectroscopy (DLTS and PICTS) analysis of different samples irradiated at increasing fluences. The defect evolution was correlated with the transport properties of the irradiated detector, always comparing with the un-irradiated one. The charge collection efficiency degradation of Schottky detectors induced by neutron irradiation was related to the increasing concentration of defects as function of the neutron fluence.

Safety, Security And safeguards In GEN IV sodium fast reactors

This work presents first a study of the national and international laws in the fields of safety, security and safeguards. The international treaties and the recommendations issued by the IAEA as well as the national regulations in force in France, the United States and Italy are analyzed. As a result of this, a comparison among them is presented. Given the interest of the Japan Atomic Energy Agency for the aspects of criminal penalties and monetary, also the Japanese case is analyzed. The main part of this work was held at the JAEA in the field of proliferation resistance (PR) and physical protection (PP) of a GEN IV sodium fast reactor. For this purpose the design of the system is completed and the PR & PP methodology is applied to obtain data usable by designers for the improvement of the system itself. Due to the presence of sensitive data, not all the details can be disclosed. The reactor site of a hypothetical and commercial sodium-cooled fast neutron nuclear reactor system (SFR) is used as the target NES for the application of the methodology. The methodology is applied to all the PR and PP scenarios: diversion, misuse and breakout; theft and sabotage. The methodology is applied to the SFR to check if this system meets the target of PR and PP as described in the GIF goal; secondly, a comparison between the SFR and a LWR is performed to evaluate if and how it would be possible to improve the PR&PP of the SFR. The comparison is implemented according to the example development target: achieving PR&PP similar or superior to domestic and international ALWR. Three main actions were performed: implement the evaluation methodology; characterize the PR&PP for the nuclear energy system; identify recommendations for system designers through the comparison.

Severe Accident Simulation of Small Modular Reactors

Since the Three Mile Island Unit 2 (TMI-2), accident in 1979 which led to the meltdown of about one half of the reactor core and to limited releases of radioactive materials to the environment, an important international effort has been made on severe accident research. The present work aims to investigate the behaviour of a Small Modular Reactor during severe accident conditions. In order to perform these analyses, a SMR has been studied for the European reference severe accident analysis code ASTEC, developed by IRSN and GRS. In the thesis will be described in detail the IRIS Small Modular Reactor; the reference reactor chosen to develop the ASTEC input deck. The IRIS model was developed in the framework of a research collaboration with the IRSN development team. In the thesis will be described systematically the creation of the ASTEC IRIS input deck: the nodalization scheme adopted, the solution used to simulate the passive safety systems and the strong interaction between the reactor vessel and the containment. The ASTEC SMR model will be tested against the RELAP-GOTHIC coupled code model, with respect to a Design Basis Accident, to evaluate the capability of the ASTEC code on reproducing correctly the behaviour of the nuclear system. Once the model has been validated, a severe accident scenario will be simulated and the obtained results along with the nuclear system response will be analysed.

Development of Core Design Methods and Tools for Gen-IV Heavy Liquid Metal Cooled Reactors

Heavy Liquid Metal Cooled Reactors are among the concepts, fostered by the GIF, as potentially able to comply with stringent safety, economical, sustainability, proliferation resistance and physical protection requirements. The increasing interest around these innovative systems has highlighted the lack of tools specifically dedicated to their core design stage. The present PhD thesis summarizes the three years effort of, partially, closing the mentioned gap, by rationally defining the role of codes in core design and by creating a development methodology for core design-oriented codes (DOCs) and its subsequent application to the most needed design areas. The covered fields are, in particular, the fuel assembly thermal-hydraulics and the fuel pin thermo-mechanics. Regarding the former, following the established methodology, the sub-channel code ANTEO+ has been conceived. Initially restricted to the forced convection regime and subsequently extended to the mixed one, ANTEO+, via a thorough validation campaign, has been demonstrated a reliable tool for design applications. Concerning the fuel pin thermo-mechanics, the will to include safety-related considerations at the outset of the pin dimensioning process, has given birth to the safety-informed DOC TEMIDE. The proposed DOC development methodology has also been applied to TEMIDE; given the complex interdependence patterns among the numerous phenomena involved in an irradiated fuel pin, to optimize the code final structure, a sensitivity analysis has been performed, in the anticipated application domain. The development methodology has also been tested in the verification and validation phases; the latter, due to the low availability of experiments truly representative of TEMIDE's application domain, has only been a preliminary attempt to test TEMIDE's capabilities in fulfilling the DOC requirements upon which it has been built. In general, the capability of the proposed development methodology for DOCs in delivering tools helping the core designer in preliminary setting the system configuration has been proven.

Smantellamento di un reattore di potenza gas-grafite di I generazione: metodologie e tecnologie di rimozione della grafite e smantellamento del nocciolo

Il presente lavoro di tesi ha riguardato lo studio dello smantellamento di un reattore gas grafite di potenza di I Gen. L’indagine è stata focalizzata in particolare al recupero della grafite irraggiata che ne costituisce il core. Viene presentata una descrizione referenziata del reattore e dei suoi componenti per mettere in evidenza la particolare architettura e le specifiche problematiche ad essa correlate. A valle di un’indagine sulle esperienze internazionali in merito al decommissioning e allo smantellamento di questi tipi di reattori, si forniscono una possibile sequenza di accesso alla cavità del reattore e una procedura per il suo smantellamento; si descrivono sommariamente le tecnologie di taglio e di handling, attualmente allo stato dell’arte, considerate come più idonee a questo tipo di applicazione. Vengono descritte le principali criticità della grafite nuclear grade ed illustrati i fenomeni caratteristici che ne determinano l’evoluzione nel reattore. Sulla base dei dati resi disponibili dalla Sogin S.p:A. e ricorrendo ai dati di letteratura per quelli non disponibili, è stato effettuato un assessment della grafite irraggiata costituente il nocciolo del reattore, rivolto in particolare a determinarne le caratteristiche meccaniche e la resistenza residua post-irraggiamento. Per valutare la possibilità di prelevare la grafite dal nocciolo è stato ipotizzato un dispositivo di presa che agganci per attrito i blocchi di grafite del moderatore attraverso il canale assiale. Infine è stata valutata la fattibilità di tale metodo attraverso una serie di simulazioni agli elementi finiti dirette a verificare la resistenza del blocco in varie condizioni di carico e vincolo. Come risultato si è dimostrata la fattibilità, almeno in via preliminare, del metodo proposto, determinando l’inviluppo di utilizzo del dispositivo di presa nonché la compatibilità del metodo proposto con le tecnologie di handling precedentemente individuate.

Hydrogen production by enhanced methane reforming with membrane reactors or with CO2 capture materials

Pure hydrogen production from methane is a multi-step process run on a large scale for economic reasons. However, hydrogen can be produced in a one-pot continuous process for small scale applications, namely Low Temperature Steam Reforming. Here, Steam Reforming is carried out in a reactor whose walls are composed by a membrane selective toward hydrogen. Pd is the most used membrane material due to its high permeability and selectivity. However, Pd deteriorates at temperatures higher than 500°C, thus the operative temperature of the reaction has to be lowered. However, the employment of a membrane reactor may allow to give high yields thanks to hydrogen removal, which shifts the reaction toward the products. Moreover, pure hydrogen is produced. This work is concentrated on the synthesis of a catalytic system and the investigation of its performances in different processes, namely oxy-reforming, steam reforming and water gas shift, to find appropriate conditions for hydrogen production in a catalytic membrane reactor. The catalyst supports were CeZr and Zr oxides synthesized by microemulsion, impregnated with different noble metals. Pt, Rh and PtRh based catalysts were tested in the oxy reforming process at 500°C, where Rh on CeZr gave the most interesting results. On the opposite, the best performances in low temperature steam reforming were obtained with Rh impregnated on Zr oxide. This catalyst was selected to perform low temperature steam reforming in a Pd membrane reactor. The hydrogen removal given by the membrane allowed to increase the methane conversion over the equilibrium of a classical fixed bed reactor thanks to an equilibrium shift effect. High hydrogen production and recoveries were also obtained, and no other compound permeated through the membrane which proved to be hydrogen selective.

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.