[en] AREVA is involved in several projects for the development of fuels with enhanced accident tolerance. Through its participation in the DOE-NE ATF programme, AREVA is investigating with the University of Florida new UO_2 pellets containing SiC additives as whiskers or particles, and fabricated with Spark Plasma Sintering technique, which reduces fabrication times. These new pellets have the potential to increase the thermal conductivity by up to 60% of the conventional UO_2 pellet, which will therefore decrease the pellet temperature during operation and thus decrease fission gas release. Nevertheless, this still has to be confirmed at very high temperatures and especially under irradiation. Concerning potential cladding solutions the coating of zirconium alloy with a MAX phase is one option that is investigated. The goal is to limit the zirconium oxidation reaction and its production of hydrogen during high temperature steam corrosion. Additionally, AREVA is also involved as fuel vendor in the development of a molybdenum (Mo) cladding managed by the Electric Power Research Institute (EPRI). The latter uses the good high temperature thermo-mechanical properties of molybdenum (high thermal conductivity and mechanical strength) to improve accidental behaviour. Furthermore, AREVA is actively involved with the CEA and EDF in tri-partite R&D projects to develop the CEA’s two potential cladding concepts of chromium coated zirconium alloys and sandwich SiC/SiC composite cladding. The chromium-coated zirconium alloys have shown great potential at inhibiting the high temperature steam oxidation reaction and preserving the cladding mechanical properties. The SiC/SiC sandwich cladding, initially developed for fast breeder reactor applications, exhibits low steam oxidation kinetics at high temperature, which would enhance the LWR fuel’s accidental behaviour. (author)

[en] The present work is part of the Fourth Generation Reactor Program, which describes the methodology and results for joining SiC substrates by metallic silicides with SiC powder reinforcements. The severest temperatures in service are in the range of 1000 C but short-time incursions at 1600 or 2000 C have to be anticipated. One of the key issues is the joining of SiC_f/SiC_m composites to seal the combustible cladding. We describe the results for joining SiC substrates in liquid state using TiSi_2. Joint integrity and joint strength can be improved by adding small SiC particles to the silicides powders. The assemblies are obtained in an inductive furnace. Cross-sections of the assembly, wettability tests, thermo-mechanical properties, and four-point bending tests are presented. (authors)

[en] Conclusions and perspectives: Designing GFR fuel elements represents a considerable challenge. There does seem to be little (if any) alternative to relying on SiC/SiC-based solutions. This resulted in CEA’s R&D program facing several technological bottlenecks. Innovative solutions were proposed and patented in the recent years: • Mixed ceramic/metal SA-duct; • “Buffer” bond; • Blind-end SiC/SiC cladding; • “Sandwich” cladding. Much progress has been accomplished and many perspectives have been open… …but proof is yet to be established that the proposed concepts are truly viable. Despite the present slowing down of GFR-dedicated R&D, limited studies are still conducted by CEA within the frame of LWR/SFR-dedicated programs

[en] SiC/SiC composite is a refractory material which presents an excellent mechanical behavior in very harsh conditions (high temperature and high irradiation flux). Initially SiC/SiC composite was developed as cladding materials for the Fourth generation Gas cooled Fast Reactor (GFR) and is now considered for use in Sodium-cooled Fast Reactors (SFR). CEA has to establish the feasibility of a SiC fiber reinforced SiC composite as structural material of the fuel element of the SFR. The behaviour of this fuel element that contains the UPuOx compound of actinides up to 50% by volume is subjected to severe operating conditions (up to 100 dpa; up to 700 C. degrees in nominal conditions and about 1600 C. degrees in fourth category accident). The cladding material must meet stringent requirements such as: fabricability, tightness and compatibility with sodium, high thermo-mechanical properties, sufficient thermal conductivity, low swelling properties and irradiation creep. We describe herein the last improvements regarding fabrication of tubular SiC/SiC composites through chemical vapour infiltration (CVI) route for generic in-core structures applications. 3 particular aspects of fabrication are considered: the fabrication of the optimized weaving, the smoothing of SiC/SiC surfaces and joining. The CVI has shown the extent of its possibilities for the development of tubes through judicious selection of fiber architecture. Compliance with the geometric tolerance on the diameter of the tube can be achieved without machining operation; in addition, it has been shown that the strain to failure exceeds the target value of 0.5%. The work of smoothing the outer surface of the first SiC/SiC ceramic matrix composites gives satisfactory results; the method proves to be industrially reliable. The sealing system of needle cap brazed poses no particular problem

[en] (A3600GV) We propose a lightweight method for the determination of heat difusivity of silica fiber bundles based on the use of a laser and an IR camera. The fiber bundle is maintained in traction in a holder; exposition is made as a step function, followed by a laser shutdown. The movie obtained by the IR camera is then processed : frame averaging, background computation and subtraction, image smoothing, extraction of the IR signal along the fiber bundle. A 1D model has been developed. This problem admits an analytic solution that we have obtained through the use of Laplace transforms. Several identification methods are proposed and tested, and have been compared favorably with an existing method based on periodic heating. Results are in agreement with literature values.

[en] Over the period 2002-2012, CEA conducted some extensive R&D on the design of GFR fuel elements (together with related material and core/system studies). This paper reviews the challenges raised by this programme, the solutions proposed to address them, and the remaining issues. Studies were performed on the assembly duct, the pin bundle and the fuel pin. The main issues were related to the challenge of using silicon carbide composites (SiC/SiC) for the pin cladding and the assembly duct, as well as mixed uranium-plutonium carbide (UPuC) for the nuclear fuel. Emphasizing the pin design, key achievements are reviewed in this paper regarding such topics as fission product confinement and high burnup performance, for the sake of which original design options were recently patented. (author)

[en] Among various refractory materials, SiC/SiC ceramic matrix composites (CMC) are of prime interest for fusion and advanced fission energy applications, due to their excellent irradiation tolerance and safety features (low activation, low tritium permeability,K). Initially developed as fuel cladding materials for the Fourth generation Gas cooled Fast Reactor (GFR), this material has been recently envisaged by CEA for different core structures of Sodium Fast Reactor (SFR) which combines fast neutrons and high temperature (500 deg.C). Regarding fuel cladding generic application, in the case of GFR, the first challenge facing this project is to demonstrate the feasibility of a fuel operating under very harsh conditions that are (i) temperatures of structures up to 700 deg.C in nominal and over 1600 deg.C in accidental conditions, (ii) irradiation damage higher than 60 dpa_SiC, (iii) neutronic transparency, which disqualifies conventional refractory metals as structural core materials, (iv) mechanical behavior that guarantees in most circumstances the integrity of the first barrier (e.g.: ε> 0.5%), which excludes monolithic ceramics and therefore encourages the development of new types of fibrous composites SiC/SiC adapted to the fast reactor conditions. No existing material being capable to match all these requirements, CEA has launched an ambitious program of development of an advanced material satisfying the specifications [1]. This project, that implies many laboratories, inside and outside CEA, has permitted to obtain a very high quality compound that meets most of the challenging requirements. We present hereinafter few recent results obtained regarding the development of the composite. One of the most relevant challenges was to make a gas-tight composite up to the ultimate rupture. Indeed, multi-cracking of the matrix is the counterpart of the damageable behavior observed in these amazing compounds. Among different solutions envisaged, an innovative one has been successful. It consists of inserting a metallic layer between two tubes of CMC [2]. The concept, illustrated in figure 1, guaranties a perfect helium tightness up to fracture of the CMC. Another challenge was to prepare a representative cladding with very strict geometrical tolerances. Revisiting the fabrication of the entire breading process has allowed to ensure a perfect geometry of the final tube. Thanks to the high quality of manufacture and the high level of purity of composite materials manufactured at CEA, few tens of CMC objects (tubes, disks and plates) have been prepared in order to be irradiated in the Russian reactor 'BOR 60'. For the first time, composite materials will be submitted to swift neutrons at very high damaging doses (up to 80 dpa SiC) between 400 and 520 C. Post irradiation examinations expected for 2015 should give reliable results on the behavior of this multi-materials component. In parallel, other basic researches are conducted to improve the properties of the CMC and round off the understanding [3, 4, 5]. Some new results allowed to extend the field of use of the CMC through an optimization of the interphase of the composite. The figure 4 shows the relative elongation of a CMC after a two hours dwell time annealing in argon at different temperatures: optimized composite can sustain very high temperature without drastic drop of its mechanical properties. (authors)

[en] This paper gives an overview of CEA studies on advanced nuclear fuel claddings for enhanced Accident Tolerant LWR Fuel in collaboration with industrial partners AREVA and EDF. Two potential solutions were investigated: chromium coated zirconium based claddings and SiC/SiC composite claddings with a metallic liner. Concerning the first solution, the optimization of chromium coatings on Zircaloy-4 substrate has been performed. Thus, it has been demonstrated that, due in particular to their slower oxidation rate, a significant additional 'grace period( can be obtained on high temperature oxidized coated claddings in comparison to the conventional uncoated ones, regarding their residual PQ (Post-Quench) ductility and their ability to survive to the final water quenching in LOCA and, to some extent, beyond LOCA conditions. Concerning the second solution, the innovative 'sandwich' SiC/SiC cladding concept is introduced. Initially designed for the next generation of nuclear reactors, it can be adapted to obtain high safety performance for LWRs in LOCA conditions. The key findings of this work highlight the low sensitivity of SiC/SiC composites under the explored steam oxidation conditions. No signification degradation of the mechanical properties of CVI-HNI SiC/SiC specimen is particularly acknowledged for relatively long duration (beyond 100 h at 1200 Celsius degrees). Despite these very positive preliminary results, significant studies and developments are still necessary to close the technology gap. Qualification for nuclear application requires substantial irradiation testing, additional characterization and the definition of design rules applicable to such a structure. The use of a SiC-based fuel cladding shows promise for the highest temperature accident conditions but remains a long term perspective

[en] Ceramic matrix composites have been identified as a potential material of core structure for the fourth generation of fission nuclear reactors. Regarding their excellent mechanical behavior in very harsh conditions (high temperature and high irradiation flux), the CVI-SiC_f/SiC composites with pyrocarbon interlayer are of prime interest for the fuel cladding in the gas-cooled fast reactor. Although the working atmosphere is helium in these advanced reactors, the presence of oxidizing impurities could have a significant role on the mechanical behavior of materials subjected to long-term exposures. Within this framework, this study was intended to investigate the influence of oxidation on the SiC_f/SiC composites mechanical properties. Different pre-damage states were intentionally introduced by mechanical tensile tests on plate specimens before performing an oxidation treatment of 1000 h at 1000 C under helium with 10 ppm of O₂. The degradation of the composite was determined from the mechanical behavior of post-exposure specimens. Results were correlated both with microstructural observations of the damage and with characterizations of the generated oxides at the surface of the composites. The most severe decline of mechanical properties occurs for the higher pre-damaged loadings. Indeed in this case, the silica formed during the oxidation of SiC is not in sufficient quantities to fill the cracks. (authors)

[en] Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO_2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the 2. Meeting on Increased Accident Tolerance of Fuels for LWRs. Content: 1 - Overview of the exchanges after the December-2012 Workshop through the discussion forum established at the OECD-NEA (S. Massara, NEA); 2 - Metrics Development for Enhanced Accident Tolerant LWR Fuels (S. Bragg-Sitton, INL); 3 - Candidate ATF Clad Technologies and Key Feasibility Issues (L. Snead, ORNL); 4 - CEA studies on nuclear fuel claddings for LWRs enhanced accident tolerant fuel. Some recent results, pending issues and prospects (J.C. Brachet, CEA); 5 - Current status on the accident tolerant fuel development in the Republic of Korea (J.Y. Park, J.H. Chang, KAERI); 6 - The current status of fuel R and D in the P.R. of China (T. Liu, CGN). Session 2: Key elements for a work programme on ATF: 7 - Beneficial characteristics of ATF (metrics) (L. Hallstadius, Westinghouse); 8 - Reactor types of interest (applicability) (L. Ott, ORNL); 9 - Impact on normal operations (N. Waeckel, EDF); 10 - Accident scenarios of interest (L. Ott, ORNL); 11 - Modelling tools (J.M. Bonnet, IRSN); 12 - Facilities and potential intermediate-term ATF development (L. Snead, ORNL); 13 - Experimental needs for assessment and validation (C. Vitanza, OECD-Halden); 14 - Development plan of Accident Tolerant Control Rod in Japan (H. Ohta, CRIEPI). Session 3: Organisations expectations and provisional contribution to a collaborative activity on ATF: 15 - Links with IAEA activity (G. Dyck). Session 4 - Collaborative framework: 16 - Examples of NEA Committees Structure (J. Gulliford, NEA)