[en] Hundred micrometer thick specimens of 9% Cr martensitic steels EM10 and T91 were homogeneously implanted with He⁴ to concentrations up to 0.5 at.% at temperatures from 150 to 550 deg. C. The specimens were tensile tested at room temperature and at the respective implantation temperatures. Subsequently the fracture surfaces were analysed by scanning electron microscopy and some of the specimens were examined in an instrumented hardness tester. The implanted helium caused hardening and embrittlement which both increased with increasing helium content and with decreasing implantation temperature. Fracture surfaces showed intergranular brittle appearance with virtually no necking at the highest implantation doses, when implanted below 250 deg. C. The present tensile results can be scaled to tensile data after irradiation in spallation sources on the basis of helium content but not on displacement damage. An interpretation of this finding by microstructural examination is given in a companion paper [J. Nucl. Mater., these Proceedings]

[en] Highly adherent, thin Cr coatings on Zr-based nuclear fuel claddings can be potentially used for the development of accident-tolerant fuels in light water reactors. To guarantee the successful implementation of Cr-coated Zr alloys as cladding tubes in nuclear power plants, the adhesive strength of the Cr coatings must be assessed. The interface between Cr and Zr was characterized via high-resolution transmission electron microscopy. We observed the formation of nanometer-thick Zr(Fe, Cr)₂ poly-type, structured Laves phases at the interfacial region that display both C14 and C15 lattice symmetries. Although the crystallinity was preserved throughout the interfacial region, different atomic configurations were observed for all the interfaces studied. In most cases, coherent or semicoherent crystallographic relationships were observed, ensuring the adhesive strength of the coating.

[en] In order to reduce the cost of electricity and to minimize the required Natural Uranium resources, increasing the cycle length and the fuel burn-up are an important issue. Specific studies have been initiated to evaluate the fuel behaviour and core loading to achieve burn-up greater than 70 GWd/t (average burn-up of the most irradiated assembly). This strategy necessarily implies an increase of the initial reactivity reserve and consequently of the capability of the control systems to compensate this reserve at the beginning of the cycle. The soluble boron quantity can be increased but must be strictly less than a maximal limit to keep the moderator coefficient at a negative value. Burnable poisons, such as Gadolinium, are usually used together with soluble boron. The neutronic properties of the Gadolinium are well known, however, in order to limit the degradation of the pin thermo-mechanical properties (especially the thermal conductivity) in the presence of this absorber, the Gd content in the pellet is generally limited to 8%. Therefore, the use of Erbium has been proposed as an alternative poison to Gadolinium. The solution proposed consists in incorporating Er into the cladding tube. The Erbium has been introduced in a low alloyed zirconium matrix as an internal liner, to avoid the degradation of the outer corrosion of the cladding due to this element in a pressurized water environment. This paper describes mostly the neutronic results obtained with such concept. Compared to other poisons, the results shows that the erbium is the best suited to be inserted in the fuel cladding tubes. To obtain the best performances, it is preferable to load a low content of erbium in all the fuel cladding tubes of the sub-assembly (homogeneous poisoning). Compared to the actual reference poisoning with few gadolinium rods, the study underlines that the core radial power distribution can be easier optimised using our concept and the cycle length can be also increased. The use of enriched erbium increases these gains. The results of fabricated fuel cladding tubes with addition of few percents of erbium are also encouraging. A French CEA patent on this concept has been registered. (authors)

[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 main objective of this paper is to summarize modelling of on-heating and on-cooling phase transformations occurring in low activation martensitic (LAM) steels. Calculations of thermodynamic equilibrium phase fractions and kinetic aspects of phase transformations have been performed by using different approaches from experimental data (CCT and TTT diagrams obtained by dilatometry). All the calculated data have been compared to an important and systematic set of experimental data obtained on different LAM steels of the 7.5-11% CrWVT a type. (orig.)

[en] In the framework of the development of generation IV nuclear reactors and fusion nuclear reactors, materials with an improved high temperature (≅650 ^oC) mechanical strength are required for specific components. The 9-12%Cr martensitic steels are candidate for these applications. Thermomechanical treatments including normalisation at elevated temperature (1150 ^oC), followed by warm-rolling in metastable austenitic phase and tempering, have been applied on the commercial Grade 91 martensitic steel in order to refine its microstructure and to improve its precipitation state. The temperature of the warm-rolling was set at 600 ^oC, and those of the tempering heat-treatment at 650 ^oC and 700 ^oC thanks to MatCalc software calculations. Microstructural observations proved that the warm-rolling and the following tempering heat-treatment lead to a finer martensitic microstructure pinned with numerous small carbide and nitride particles. The hardness values of thermomechanically treated Grade 91 steel are higher than those of the as-received Grade 91. It is also shown that the yield stress and the ductility of the thermomechanically treated Grade 91 steel are significantly improved compared to the as-received material. Preliminary creep results showed that these thermomechanical treatments improve the creep lifetime by at least a factor 14.

[en] Chromium-coated zirconium alloys are being studied as Enhanced Accident Tolerant Fuel Cladding for Light Water Reactors (LWRs). Those materials are especially studied to improve the oxidation resistance of LWRs current fuel claddings in nominal and at High Temperature (HT) for hypothetical accidental conditions such as LOss of Coolant Accident. Beyond their HT behavior, it is essential to assess the materials behavior under irradiation. A first generation chromium/Zircaloy-4 interface was thus irradiated with 20 MeV Kr⁸⁺ ions at 400 °C up to 10 dpa. High-Resolution Transmission Electron Microscopy and chemical analysis (EDS) were conducted at the Cr/Zr interface. The atomic structure of the interface reveals the presence of Zr(Fe, Cr)₂ Laves phase, displaying both C14 and C15 structure. After irradiation, only the C14 structure was observed and atomic row matching was preserved across the different interfaces, thus ensuring a good adhesion of the coating after irradiation.

[en] Cr-coated M5(Framatome) cladding materials are studied and developed within the CEA-Framatome-EDF French nuclear fuel joint program as Enhanced Accident Tolerant Fuel claddings for Light Water Reactors. The objective of this paper is to bring some insights into the relationship between Equivalent Cladding Reacted (ECR) parameters, oxygen diffusion/partitioning and Post-Quench (PQ) ductility of Cr-coated M5(Framatome) fuel claddings oxidized in steam at 1200 degrees C. The physical meaning of the ECR parameter, evaluated experimentally from the measured Weight Gain (WG) or calculated using time and temperature correlations such as the Baker-Just (BJ) or Cathcart-Pawel (CP) kinetics correlations, is discussed in the light of the benefit brought by Cr coating to oxidation resistance of cladding. As shown in this article, when applied to the Cr-coated M5(Framatome) materials, the 'experimental' ECR derived from WG does not have the same physical meaning than for the uncoated cladding materials. As discussed in the paper, this is fundamentally due to the use of the ECR as a surrogate for retained ductility for uncoated claddings, and to the differences between uncoated and Cr-coated cladding in the high temperature (HT) steam oxidation processes and partitioning of the oxygen between the different layers of the oxidized cladding. It is shown in this article that Cr-coated M5(Framatome) cladding brings significant additional time-at-temperature before full embrittlement of the cladding after one-sided oxidation at 1200 degrees C and quenching, compared to uncoated materials. The oxidation times and associated Baker-Just ECR (BJ-ECR) values, above which the cladding becomes brittle after low temperature quenching, are respectively ten times and three times higher than the ones for the uncoated reference cladding. When analyzing the PQ ductility of the Cr-coated M5(Framatome) cladding using a similar methodology as the one used to derive the ECR criterion for uncoated cladding, the 1-2% ductility limit corresponds to a BJ-ECR of about 50% or higher, for a 12-15 mm-thick Cr-coated cladding tested herein. (authors)