The LCMl (Laboratorv for characterization of irradiated materials), located in CEA from Saclav. is in charge oi the mechanical tests on iriaiated materials. The dynamic tensiie testing machine, in a hot celi equipped with two remote handlings. has been first improved in 1995. to fulfiil the French safety programs on Reactivity Initiated Accident C A 0 1 .::U SCI .5 <l :11 S ~ 2 . n rr i to oni.3 11 rwt;ua,~cs cropcn, cnla on r..rrrrl L ,:a", c m " ng $,p4 ' l r U e u l i q . m t t , in,: -;u~ o j s ieapmsc .noer H A or .OCA lrios ert cscnq 3n.1 lnc!rnn coroton; Fc' lnt HIA. 1.1 2 illem, 1 2 i l ?g 1 wdr8 r.,'e, .p 1, 5 i' a m nsal ng rd!es .p !o 210.C S . ;n F 1% _OS? 01 CS dnl ACC odi i s . OCA led g ;: sra r i.?lr, U. 10 an0 n c x r g rales u. ?WC s KO. o oi. a00102r .?h: The tensile samples are machined with a spark erosion machine. directiy from pieces of cladding previously defueied. Two kinds of samples can be machined in the ciadding: Axial samples in order to test axial mechanical characteristics Rina samoles in order to test transverse mechanical characteristics, more reoresentative of HIA On one hand. the axial tensile tests were performed using the Jouie effect, and heating rates up to about 500°C.S' were obtained. This enabled us to perform the axiai tests in a satisfactory manner. On the other hand, the tensiie ring tests were first performed in a vertical furnace with a heating rate about 09T .s ' and a thermal stability about 1°C. For temperatures above 480°C. the mechanical characteristics showed a sharp drop which could be attributed to irradiation defect annealing. Therefore we have recently deveiaped an Induction heating system to reach heating rates high enough (200'c.s~') to prevent any significant anneaiing before performing the ring tensile tests. To apply a uniaxiai tangential tension, two matching half-cylinders are inserted inside the ring and are pulled apait. The main objective of this paper is to present this system that can be telemanipulated and achieve heatlng rates up to ~ O O ~ C . ~ . ' while taking into account the requirement forair-cooled coils in the hat celi. The same interface is 40 Aver@ ei ar /lnduclion heating on dynamic lensile rests in CEA Saclay used for mduction heating equipment and Joule effect (current system), in order to control the specimen temperature synchronously with the Iaadlextension, colleCting data from all transducers connected including load/displacement.

Next generation nuclear systems will require structural (including cladding) materials capable of withstanding normal operating and accidental conditions drastically more demanding than those typically encountered in current reactors. The main requirements for the structural materials to be used in these reactor systems are dimensional stability under irradiation, acceptable evolution under ageing of the mechanical properties and a good behaviour in corrosive environments. These requirements have to be met not only under normal operating conditions, but also in off normal and accident conditions.These challenging requirements imply that in most cases, the use of nuclear materials used in the current reactors is excluded, therefore a new range of materials has to be developed and qualified for nuclear use. In this paper, an overview of various materials that are essential to establish advanced system feasibility and performance is discussed as well as the various natures of R&D needed to select, optimize, assess and qualify advanced materials. (C) 2015 Elsevier B.V. All rights reserved.

The performance of hydrogen production via thermochemical cycles is typically evaluated using thermal efficiency. In this study, the sulfur-iodine cycle with heat supplied by a high-temperature reactor (HTR) is analyzed. Two cases are examined: one flow sheet designed by General Atomics in the United States, the other by Commissariat a l'energie atomique et aux energies alternatives in France. In each case, HTR helium inlet and outlet temperatures are specified. Differences in these temperature specifications lead to process variations between the limy sheets and in how the hydrogen processes interface with the nuclear heat source. Two principal conclusions result from the analysis. First, the thermal efficiency tends to plateau above a certain outlet helium temperature. This is a characteristic effect of the method of Ozturk et al. for sulfuric acid decomposition. Second, it is clear that it is impractical to discuss efficiencies for the hydrogen process that are independent of defined operating parameters of the HTR.

It is already 10 years since the (European) High Temperature Reactor Technology Network (HTR-TN) launched a program for development of HTR technology, which expanded through three successive Euratom framework programs, with many projects in line with the network strategy. Widely relying in the beginning on the legacy of the former European HTR developments (DRAGON, AVR, THTR, etc.) that it contributed to safeguard, this program led to advances in HTR/VHTR technologies and produced significant results, which can contribute to the international cooperation through Euratom involvement in the Generation IV International Forum (GIF). the main achievements of the European program, performed in complement to efforts made in several European countries and other GIF partners, are presented: they concern the validation of computer codes (reactor physics, as well as system transient analysis from normal operation to air ingress accident and fuel performance in normal and accident conditions), materials (metallic materials for vessel, direct cycle turbines and intermediate heat exchanger, graphite, etc.), component development, fuel manufacturing and irradiation behavior, and specific HTR waste management (fuel and graphite). Key experiments have been performed or are still ongoing, like irradiation of graphite and of fuel material (PYCASSO experiment), high burn-up fuel PIE, safety test and isotopic analysis, IHX mock-up thermohydraulic test in helium atmosphere, air ingress experiment for a block type core, etc. Now HTR-TN partners consider that it is time for Europe to go a step forward toward industrial demonstration. In line with the orientations of the “Strategic Energy Technology Plan (SET-Plan)” recently issued by the European Commission that promotes a strategy for development of low-carbon energy technologies and mentions Generation IV nuclear systems as part of key technologies, HTR-TN proposes to launch a program for extending the contribution of nuclear energy to industrial process heat applications addressing (1) the development of a flexible HTR that can be coupled to many different process heat and cogeneration applications with very versatile requirements, (2) the development of coupling technologies for such coupling, (3) the possible adaptations of process heat applications required for nuclear coupling, and (4) the integration and optimization of the whole coupled system. As a preliminary step for this ambitious program, HTR-TN endeavors to create a strategic partnership between nuclear industry and R&D and process heat user industries.

Gas-cooled reactors take up a strong second role in France's R&D strategy on future nuclear energy systems as priority was given in 2005 to fast neutron reactors with multiple-recycle for their potential to optimally use uranium resource and minimize the long term burden of radioactive waste. Owing to the European past experience on sodium-cooled fast reactors (SFRs), this reactor type was logically selected as reference for a new generation fast neutron reactor intended to be tested as a prototype in the 2020s and be ready for industrial deployment around 2040. At the same time, the potential merits of a gas fast reactor (GFR) with ceramic clad fuel for a safe management of cooling accident are acknowledged for the potential of this reactor type to resolve critical issues of liquid fast reactors (safety, operability and reparability). A pre-feasibility report on a first concept of GFR was issued in 2007 that summed-up results of a 5-year international R&D effort on GFR fuel technology, reactor design and operating transient analyses. This report established a global confidence in the feasibility of this concept and its potential for attractive performances. Furthermore, it suggested directions of R&D to generate by 2012 an updated concept with improved performances and taking better benefit from GFR specific technologies. A second activity on gas-cooled reactors originates from the current interest of CEA's industrial partner AREVA in high or very high temperature reactors (V/HTR) for supplying hydrogen, synthetic hydrocarbon fuels and process heat for the industry. This activity currently encompasses R&D on V/HTR key technologies such as particle fuel fabrication, high temperature compact heat exchangers and coupling technologies to various power conversion systems. R&D on V/HTR and GFR are synergistic in various respects. The GFR can be viewed as a more sustainable version of the VHTR and synergies exist in research on heat resisting materials, helium system technology and power conversion systems. Both reactors require active research in materials and spur developments of new metallic alloys and ceramics applicable to other advanced nuclear systems.

Due to its high operating temperature (up to 850°C with present technologies, possibly higher in the longer term), and its power range (a few hundred MW), the modular HTR could address a larger scope of industrial process heat needs than other present nuclear systems. Even if HTR can contribute to competitive electricity generation, this potential for industrial heat applications is the main incentive for developing this type of reactor, as it could open to nuclear energy a large non-electricity market. However several issues must be addressed and solved successfully for HTR to actually enter the market of industrial process heat: 1) as an absolute prerequisite, to develop a strategic alliance of nuclear industry and R&D with process heat user industries. 2) to solve some key technical issues, as for instance the design of a reactor and of a coupling system flexible enough to reconcile a single reactor design with multiple applications and versatile requirements for the heat source, and the development of special adaptations of the application processes or even of new processes to fit with the assets and constraints of HTR heat supply, 3) to solve critical industrial issues such as economic competitiveness, availability and 4) to address the licensing issues raised by the conjunction of nuclear and industrial risks. In line with IAEA initiatives for supporting non-electric applications of nuclear energy and with the orientations of the SET-Plan of the European Commission, the (European) HTR Technology Network (HTR-TN) proposes a new project, together with industrial process heat user partners, to provide a first impetus to the strategic alliance between nuclear and non-nuclear industries. End user requirements will be expressed systematically on the basis of inputs from industrial partners on various types of process heat applications. These requirements will be confronted with the capabilities of the HTR heat source, in order to point out possible discrepancies and issues, to assess the feasibility of different coupling schemes and to identify development needs. Partners from nuclear regulatory organisations will also address the feasibility of licensing such coupling schemes. The issues they will raise will be taken into consideration for defining coupling design bases and identifying R&D needs. A detailed roadmap for designing an industrial demonstrator of a HTR coupled with process heat applications will be inferred from this analysis, as well as R&D actions required for supporting the development of the reactor, of the coupling system and of possible adaptations or innovations in industrial processes.

A R&D programme has been launched addressing the needs of the development of an indirect cycle flexible modular HTR operating at 850°C for electricity generation and/or heat production for industrial processes. In the frame of this program, several significant technical challenges required to demonstrate the viability and performance of the system have been successfully addressed. Design and safety analysis needed the development of computational tools, therefore reactor physics, and thermo-fluid dynamics codes have been developed and are now in the process of being validated in the frame of international code-to-code and code to experiment benchmarks. Most importantly, the performance of the HTR/VHTR fuel identified as TRISO-coated particle must prove to be excellent in operating as well as accidental conditions. A manufacturing and quality control process has been developed and now fuel qualification based on irradiation and heating safety tests is being prepared on the basis of irradiation programs in France and in the frame of the GENERATION IV International Forum (GIF) as well as the development of fuel behaviour models including performance data, failure particle prediction and long-term integrity of the coating. Material and component technologies have been investigated in normal and accident conditions for V/HTR objectives. Significant progress has been made for vessel structures and reactor core structural elements. Major challenges still lie ahead for plate type compact intermediate heat exchangers, especially at temperatures above 850°C, but an alternative solution with helical tubes is also being developed. In order to demonstrate that materials have adequate performance over long service life under impure helium environment and constraints, the research programme focuses on microstructural and mechanical property data, long-term irradiation behaviour, corrosion, modelling and codification of design rules as well as qualification of components in representative helium test loops. The potential of this type of reactor for higher performances in terms of fuel burn-up and temperature (VHTR objective) has been explored, in particular for application to hydrogen production. The major research axes on hydrogen production technologies include the development and optimization of high temperature electrolysis and thermo-chemical water splitting processes such as sulphur/iodine or hybrid sulphur. Alternative thermo-chemical hydrogen generation processes operating at lower temperatures are also investigated. This paper addresses the R&D work performed since 2001 and the future work anticipated until 2012, where decisions about a demonstrator could be made at a European level within the Sustainable Nuclear Energy Technological Platform (SNE-TP). This program is strongly connected to the Euratom Framework Programmes as well as to GIF.

Le Comité de l'énergie atomique de mars 2005 a retenu le développement en parallèle de deux filières de réacteurs à neutrons rapides (RNR) avec cycle du combustible fermé. Ces deux filières doivent être étudiées pour une industrialisation à l'horizon 2040, avec un choix entre les deux filières en 2015 : Sodium Fast Reactor (SFR) et Gas Fast Reactor (GFR). Pour préparer ce choix, le CEA a lancé, avec ses partenaires EDF et AREVA, un important programme de R&D pour valider les choix de matériaux pour les structures fixes et pour le galnage combustible, en cohérence avec les sollicitations auxquelles ils seront soumis, et pour préparer la codification associée. Des moyens importants sont mis en oeuvre pour faire face aux enjeux scientifiques, à la fois au niveau expérimental, au niveau de la modélisation et de la simulation, et au niveau d'une organisation qui privilégie la collaboration : entre États, entre organisme de recherche, exploitant et constructeur, et entre disciplines scientifiques : mécanique, matériaux, corrosion.

The PROMETRA material testing program is a support program related to the study of high burnup fuel rod behavior under Reactivity Initiated Accidents (RIA) and to the interpretation of the CABRI REP-Na RIA test results. Hoop and axial tensile tests have been performed on fresh and irradiated Zircaloy-4 cladding alloy first at CEA Grenoble hot labs and now at CEA Saclay in order to assess the cladding mechanical behavior during RIA transients. Efforts have been continuously carried out in order to improve the prototipicallity of the tests for RIA studies involving new specimens and new testing techniques. The corrosion level of irradiated specimens reached up to 130 μm of oxide layer thickness. The influence of in-pile oxide layer spallation has also been addressed. High strain-rate material properties of irradiated Zircaloy-4 and the consequences of hydride embrittlement can be derived from the PROMETRA program.