[en] LFR share the main issues of all Fast Reactors, while presenting specific issues due to the use of lead as coolant. A number of constraints impairs the design of a LFR core, possibly resulting in a viability domain not exploitable for producing electricity in an efficient (hence economic) way. In particular, the most restrictive issues to be faced pend on the cladding. The selection of proper cladding materials provides the solution for the issues impairing the resistance of the cladding against stresses and irradiation effects. On the other hand, the protection of the cladding requires surface protections like oxide scales (passivation) or adherent layers (coating). Oxide scales seem not sufficient for a stable and effective protection of the base material. The application of adherent layers seems the only promising solution for protecting the cladding against corrosion. For the short term (i.e.: ALFRED), advanced 15/15Ti with coating is the reference solution for the cladding, allowing a core design complying with all the design constraints and goals. The candidate coatings are already being tested under irradiation to proceed towards qualification. In parallel, new base materials and/or coatings are presently under investigation. For the long term (i.e.: ELFR), the availability of such advanced materials/coatings might allow the extension of the viability domain towards higher and broader ranges (temperature, dpa, etc.), extending the fields of applications of LFRs and resulting in higher performances

[en] The 2010-2012 implementation plan of the European Sustainable Nuclear Industrial Initiative (ESNII), prepared in the frame of the Sustainable Nuclear Energy Technology Platform (SNETP), establishes a very tight time schedule for the start of construction of the European Gen IV prototypes; namely the construction of the LFR ETPP (European Technology Pilot Plant) MYRRHA will start in 2014 and that of the SFR Prototype ASTRID will start in 2017. The GEN IV reactors pose new challenges to the designers and scientists in terms of higher operating temperature, higher fuel burn-up, and in some cases more corrosive environment with respect to the present technologies and which impacts the materials performance. In this frame, the MATTER (Materials Testing and Rules) Project starts well targeted R and D activities to perform careful materials studies in GEN IV operational conditions and to find out criteria for the correct use of these materials in relevant reactor applications. Aim of the MATTER Project (that involved 27 partners and will end in 2015) is to complement the materials researches, in the frame of the European Energy Research Alliance (EERA) guidelines, with the implementation of pre-normative rules. The MATTER Project is divided in 3 technical Domains (called DM): DM1 - Development of test and evaluation guidelines for structural materials: to develop/establish best practice guidelines for testing and evaluation procedures, which are aimed to screen and characterize nuclear materials for innovative nuclear systems. DM2 - Pre-normative R and D for Codes and Standards: Pre-normative activities are performed, comprehensive of experiments, to revise and update the design rules (with an EU level consensus) in order to answer to some short term needs of the two projects ASTRID and MYRRHA with respect to the design and the construction of structural components. DM3 - Joint Program Scheme, implementation and Priorities: to optimise the effectiveness and efficiency of the EERA Joint Program on nuclear materials for innovative reactors and to support specific research activities related to fundamental understanding of ODS steels fabrication. ODS steels are considered candidate materials, in the medium-long term, for high fuel burn-up cladding application. After a brief presentation of DM1 and DM3, this paper mainly focuses on description of Pre-normative R and D activities for Codes and Standards (DM2). (authors)

[en] GEMMA EU Project addressed materials development and testing for Sodium, Heavy Liquid Metals and High Temperature Helium to be employed in GEN IV reactors. The main areas addressed are: (1) protective coatings and advanced alumina forming corrosion-resistant austenitic steels for application in heavy-liquid metal-cooled systems; (2) testing of structural steels, protective coatings as well as welds of different type under representative conditions in contact with heavy liquid metals and helium; (3) development of physical models for the prediction of the behaviour of austenitic alloys under long term irradiation. The poster describes some relevant and interesting new results and observations. (authors)

[en] Highlights: → ELSY, the European Lead Fast Reactor (LFR) design is presented. → Presentation of Main Components design. → Core design, safety systems and safety analysis. → Future development activities for Lead-cooled system. - Abstract: The conceptual design of the European Lead Fast Reactor is being developed starting from September 2006, in the frame of the EU-FP6-ELSY project. The ELSY (European Lead-cooled System) reference design is a 600 MWe pool-type reactor cooled by pure lead. The ELSY project demonstrates the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features, while fully complying with the Generation IV goal of sustainability and minor actinide (MA) burning capability. Sustainability was a leading criterion for option selection for core design, focusing on the demonstration of the potential to be self sustaining in plutonium and to burn its own generated MAs. To this end, different core configurations have been studied. Economics was a leading criterion for primary system design and plant layout. The use of a compact and simple primary circuit with the additional objective that all internal components be removable, are among the reactor features intended to assure competitive electric energy generation and long-term investment protection. Low capital cost and construction time are pursued through simplicity and compactness of the reactor building (reduced footprint and height). The reduced plant footprint is one of the benefits coming from the elimination of the Intermediate Cooling System, the low reactor building height is the result of the design approach which foresees the adoption of short-height components and two innovative Decay Heat Removal (DHR) systems. Among the critical issues, the impact of the large mass of lead has been carefully analyzed; it has been demonstrated that the high density of lead can be mitigated by compact solutions and adoption of seismic isolators. Safety has been one of the major focuses all over the ELSY development. In addition to the inherent safety advantages of lead coolant (high boiling point and no exothermic reactions with air or water) a high safety grade of the overall system has been reached. In fact the overall primary system has been conceived in order to minimize pressure drops and, as a consequence, to allow decay heat removal by natural circulation. Moreover two redundant, diverse and passive operated DHR systems have been developed and adopted. The paper presents the overall work performed so far.

[en] The goal of the ORIENT-NM action is to produce a single European strategic vision on research and innovation concerning nuclear materials in the EU, serving all reactor generations and nuclear systems. The key in this endeavour is to focus on advanced materials science practices that, combined with digital techniques, will enable acceleration in materials development, manufacturing, supply, qualification, and monitoring, in support of nuclear energy safety, efficiency, economy and sustainability. The research agenda will be rooted in existing virtuous examples of nuclear materials science projects. Here the results of three of them are summarised, thereby covering different reactor applications and families of materials, as well as a range of advanced material research approaches. GEMMA addressed a number of key areas concerning the development and qualification of metallic structural materials for GenIV reactor conditions, focusing on austenitic steels and their compatibility with several non-aqueous coolants, their welds and the modelling of their stability under irradiation. INSPYRE was an integrated project applying a basic science approach to (U,Pu)O₂ fuels, to develop physics-based models for the behaviour of nuclear fuels under irradiation and improve fuel performance codes. Modelling was also the focus of the M4F project, which brought together the fission and fusion materials communities to study the effects of localised deformation under irradiation in ferritic/martensitic steels and to develop good practices to use ion irradiation as a tool to evaluate radiation effects on materials. (authors)

[en] Highlights: ► The HETS (high-efficiency thermal shield) concept, initially developed by ENEA for water, has been adapted for use with He as coolant. ► This DEMO divertor concept is based on elements joined in series and protected by a hemispheric dome. ► It has been calculated to be capable of sustaining an incident heat flux of 10 MW/m² when operating at 10 MPa, an inlet He temperature of 600 °C, and an outlet temperature of 800 °C. ► The activity is focused on the manufacturing of a single HETS module with W armor and on its thermal–hydraulic testing. ► A CFD analysis by ANSYS-CFX was performed in order to predict the thermal–mechanical behavior of the module and a final comparison with the experimental data is required to validate the CFD results. - Abstract: The development of a divertor concept for fusion power plants that is able to grant efficient recovery and conversion of the considerable fraction (∼15%) of the total fusion thermal power incident is deemed to be an urgent task to meet in the EU Fast Track scenario. The He-cooled conceptual divertor design is one of the possible candidates. Helium cooling offers several advantages including chemical and neutronic inertness and the ability to operate at higher temperatures and lower pressures than those required for water cooling. The HETS (high-efficiency thermal shield) concept, initially developed by ENEA for water, has been adapted for use with He as coolant. This DEMO divertor concept is based on elements joined in series and protected by a hemispheric dome; it allows an increase of thermal exchange coefficient both for high speed of gas and for “jet impingement” effects of gas coming out from the internal side of hemispheric dome. It has been calculated to be capable of sustaining an incident heat flux of 10 MW/m² when operating at 10 MPa, an inlet He temperature of 600 °C, and an outlet temperature of 800 °C. The presented activity, performed in the frame of EFDA-TW5TRP-001 task, was focused on the manufacturing of a single HETS module and on its thermal–hydraulic testing. The materials used for the HETS module manufacturing were all DEMO-compatible: W for the armor material and the hemispherical-dome, DENSIMET for the exchanger body. The testing is performed by connecting the module to HEFUS3 He loop system that is a facility able to supply the He flow to the required testing conditions: 400 °C, 4–8 MPa and 20–40 g/s. The needed incident heat flux is obtained by RF inducting equipment coupled to an inductor coil installed just over the HETS module. A CFD analysis by ANSYS-CFX was performed in order to predict the thermal–mechanical behavior of the module and a final comparison with the experimental data is required to validate the CFD results. All parameters are monitored and recorded by data acquisition system.

[en] The conceptual design of the European Lead Fast Reactor is being developed starting from September 2006, in the frame of the ELSY project. The ELSY reference design is a 600 MWe pool-type reactor cooled by pure lead. The ELSY project demonstrates the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features, whilst fully complying with the Generation IV goal of sustainability and minor actinide (MA) burning capability. Sustainability was a leading criterion for option selection for core design, focusing on the demonstration of the potential to be self sustaining in plutonium and to burn its own generated MAs. To this end, different core configurations have been studied. Economics was a leading criterion for primary system design and plant layout. The use of a compact and simple primary circuit with the additional objective that all internal components be removable, are among the reactor features intended to assure competitive electric energy generation and long-term investment protection. Low capital cost and construction time are pursued through simplicity and compactness of the reactor building (reduced footprint and height). The reduced plant footprint is one of the benefits coming from the elimination of the Intermediate Cooling System, the low reactor building height is the result of the design approach which foresees the adoption of short-height components and two innovative DHR systems. Among the critical issues, the impact of the large mass of lead has been carefully analyzed; it has been demonstrated that the high density of lead can be mitigated by compact solutions and adoption of seismic isolators. Safety has been one of the major focuses all over the ELSY development. In addition to the inherent safety advantages of lead coolant (high boiling point and no exothermic reactions with air or water) a high safety grade of the overall system has been reached. In fact the overall primary system has been conceived in order to minimize pressure drops and, as a consequence, to allow decay heat removal by natural circulation. Moreover two redundant, diverse and passive operated DHR systems have been developed and adopted. The ELSY project was organised into 6 technical plus one coordination work packages. The paper presents the overall work performed so far in the different areas. (author)

[en] The European Lead Fast Reactor has been developed in the frame of the European lead system (ELSY) project funded by the Sixth Framework Programme of EURATOM. The project, coordinated by Ansaldo Nucleare, involved a wide consortium of European organizations. The ELSY reference design is a 600 MWe pool-type reactor cooled by pure lead. The project demonstrates the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features, whilst fully complying with the Generation IV goals. The paper focuses on the main aspects of the proposed design for the European lead fast reactor highlighting the innovation of this reactor concept and overall objectives. Special attention has been dedicated to safety starting from the first step of the design development taking into account other important aspects, such as the investment protection, the compactness of the primary system as well as sustainability. The main safety features of the proposed innovative decay heat removal (DHR) systems are presented. From the beginning of 2010, and for a duration of three years, the European Commission (EC) is financing the new project Lead European Advanced Demonstration Reactor (LEADER) as part of the 7th Framework Program. This paper highlights the main objectives of the LEADER project. (author)

[en] The ORIENT-NM project is elaborating a single European strategic research and innovation agenda (SRIA) that should set the path for future activities on nuclear materials in the EU, until 2040, serving all reactor generations. The key in this endeavour is to focus on advanced materials science practices that, combined with digital techniques, will enable acceleration in materials development, manufacturing, supply, qualification, and monitoring, in support of nuclear energy safety, efficiency, economy and sustainability. This research agenda will not come out of the blue: it will be rooted in existing virtuous examples of materials science projects that target nuclear energy innovation. Here three of them are considered, that cover different reactor generation applications. NUCOBAM aims at developing the qualification process and provide the evaluation of the in-service behavior of additively manufactured components in nuclear installations, as a promising technique to tackle obsolescence challenges in operating reactors and manufacture new components with optimized design, for increased safety and efficiency. GEMMA addresses a number of key areas concerning materials development and qualification for Gen IV reactor conditions, namely: corrosion-resistant austenitic steels for application in heavy-liquid metal-cooled systems; production of welds on available austenitic steels and their characterization in terms of internal stresses; testing of all these materials (baseline, welds and advanced) under representative conditions in contact with heavy liquid metals and helium; development of physical models for the prediction of the behaviour of austenitic alloys under long term irradiation. Finally, the M4F project creates a bridge between fission and fusion materials communities, by applying physical modelling techniques to target two objectives: understand and predict the origin and effects of localised deformation under irradiation in ferritic/martensitic steels affecting the mechanical behaviour of components for future fission and fusion reactors, so as to enable their design based on robust standards; to develop good practices to use ion irradiation as a tool to evaluate radiation effects on materials, also applied to ferritic-martensitic alloys. This paper will report on the key ideas of the ORIENT-NM SRIA and will highlight selected results of the NUCOBAM, GEMMA and M4F projects. (authors)

[en] In the framework of fusion materials research, a neutron source has been considered a key installation to support EU plan toward DEMO reactor design. IFMIF facility being the present proposal, a pragmatic approach to EU fusion roadmap timeline considers complementary solutions mandatory, within a shared strategy. New Sorgentina Fusion Source (NSFS) has been recently proposed in order to populate an engineering database through materials irradiation tests. Proven technology of D–T neutron generators is implemented together with ion source and accelerator devices currently used in neutral injection systems at experimental tokamaks. Deuterium and tritium enriched hydride is on-line reloaded by impinging D–T beams via ion implantation onto a high-speed rotating target – D–T retention is allowed through temperature control. Hydride metal layer is re-deposited increasing plant availability factor. Target design is proposed to cope with thermal transients and mechanical loads. Solutions to thermal fatigue concerns are presented. Irradiation capability is then enhanced attaining relevant materials exposure. Main facility characteristics are provided as well as thermal and mechanical issues.