[en] The mechanical properties of spallation target components exposed to combined effects of proton and neutron irradiations and in contact with liquid metal provide important information for the assessment of structural component integrity, which is crucial for the design of accelerator driven reactor concepts such as the MYRRHA reactor. In this study we perform tensile tests on T91 steel samples extracted from the MEGAPIE, and from the TWIN-ASTIR experiment. The tests are performed at different temperatures as well as with and without the contact with liquid metal. In both groups of samples we observed significant influence of liquid metal on the tensile properties, in particular reduction of total elongation. The influence of different conditions in two irradiation programs on the mechanical properties, such as irradiation temperature fluctuations, the presence of neutron/proton irradiation, with and without the contact with lead-bismuth eutectic, different flux and fluence, are also discussed.

No abstract available

[en] The MYRRHA reactor (Multi-purpose hYbrid Research Reactor for High-tech Applications), currently developed at SCK•CEN, will allow the demonstration of transmutation of high-level nuclear waste, fuel developments for innovative reactor systems, material developments for GEN IV and fusion reactors, and radioisotope production for medical and industrial applications. Since MYRRHA is based on heavy liquid metal technology with Lead Bismuth Eutectic (LBE) coolant, it can significantly contribute, during its development and in its operational phase, to the development of Lead Fast Reactor (LFR) technology for both large and SMR systems cooled with Lead Bismuth Eutectic or with Lead. To support the MYRRHA development, SCK•CEN has launched a strong R&D programme to address the main design and licensing challenges, in particular those related to the use of liquid Lead-Bismuth Eutectic as reactor coolant. In this frame SCK•CEN has constructed and commissioned various LBE test facilities for heavy liquid metal chemistry and conditioning research, the heavy liquid metal corrosion research for materials for advanced fast reactors, the testing of mechanical rotating components in heavy liquid metals, reactor component testing in a heavy liquid metal loop and a facility for the validation of complex flows in liquid metal pool systems. These facilities are used for the qualification of the key materials and components of MYRRHA and can also be used for the development of materials and components for fast reactors of all power ranges, including SMR type, working with LBE or Lead as coolant. The paper describes the SCK•CEN concept roadmap for lead SMR type power reactors based on MYRRHA technology developed from the ongoing R&D programme. The existing research facilities and their applicability for the development of lead SMR type systems are presented. (author)

[en] This Report is a summary of all activities at UCB and LANL related to the SMoRE program organized by the IAEA. In the following chapters results of the ion beam irradiations conducted during the participation of LANL/UCB on the program are presented. This report is organized in different chapters according to the different activities started during the SMoRE program. The focus of the UCB/LANL work is to perform ion beam irradiation on candidate structural materials, as well as basic simple materials to understand the effects of ion beam irradiation leading to a path-forward to use these techniques for accelerated materials testing and “pre-screening” of materials which can be used in nuclear applications. The work presented here is centered around small scale mechanical testing after irradiation to gain property changes on materials due to ion beam irradiation needed to compare to neutron irradiation data. Also, it allows investigation to what extent these novel techniques can be useful to study large neutron irradiated materials so that less radioactive material need to be handled while gaining additional data to enhance the statistical value. As a result of these activities, UCB/LANL started several national and international collaborations on this topic since SMoRE meetings led to new contacts in the community. In the last chapter these new collaborations are discussed based on the proposals submitted to the funding agencies.

[en] This SMoRE CRP has used a wide variety of microanalytical tools to extract data from ion beam simulations of neutron damage to reactor structural materials. While some of these tools have been used for many years, others are very new and are undergoing continuous evolution and refinement. These tools have significant advantages but also have unique limitations and disadvantages. This chapter reviews the major techniques and their characteristics, especially those that require reporting significantly more experimental details than might usually be included in general publications. (author)

[en] The Fuel Cycle R and D (FCRD) initiative is investigating methods of burning minor actinides in a transmutation fuel. To achieve this goal, the fast reactor core materials (cladding and duct) must be able to withstand very high doses (>200 dpa design goal). While mechanical testing on large samples delivers direct engineering data these types of tests are only possible if enough sample material and required hot cell capabilities are available. Small scale materials testing methods in addition to large scale materials testing allows one to gain more insight on the same specimen and directly probe areas of interest which are not accessible otherwise (small welds, areas with different microstructure, etc.). However, in order to use small scale testing techniques and to probe materials microstructures using these methods, the relationship between the different scales needs to be investigated. In order to establish a research based relationship between small scale and large scale materials testing several different mechanical testing techniques were conducted on the same specimen irradiated in the Spallation Target Irradiation Program (STIP) at the Swiss spallation source (SINQ) at the Paul Scherrer Institute (PSI) up to a dose of 19 dpa. Tensile testing, micro hardness testing and micro compression testing on focused ion beam (FIB) manufactured pillars were performed on remaining parts of tensile test specimens tested and irradiated at PSI. It is shown that the yield strength increases measured by tensile testing, micro compression testing and micro hardness testing all show the same trend. In addition FIB based techniques also allow one to cut local electrode atom probe (LEAP) samples. In this procedure samples are cut of such a small size that no radioactivity on the prepared sample can be measured. In order to establish trust in these FIB based techniques estimations of residual activity on these samples is essential and presented here. (author)

[en] Copper and phosphorus as impurities and nickel as alloying element were already a long time ago identified to significantly affect radiation embrittlement of reactor pressure vessel materials. As a result, the Cu-content in RPV materials is limited to less than 0.1 wt%, the P-content to less than 0.01 wt% and the Ni-content to less than ∼1 wt%. These three elements are explicitly considered in most of the available embrittlement trend curves. Beside Cu, P and Ni, also Mn and Si were also proposed to influence irradiation embrittlement. Indeed, these two elements are found in the irradiation-induced solute clusters examined with atom probe tomography. This paper aims to examine the possible synergy between copper and phosphorus on one hand, and nickel and manganese on the other hand. Therefore, chemically-tailored steels including twelve steels with variable copper/phosphorus contents and nine steels with variable nickel/manganese contents were irradiated at 290 °C in the BR2 high flux reactor to various fluence levels ranging between about 1 and 6 10¹⁹ n/cm². The concentrations of the key elements were selected such as to have three distinct levels covering low, medium and high levels. Finally, two steels with all Cu/P/Ni/Mn contents at their minimum and maximum levels were also selected for comparison. Typically, for each steel, small size tensile specimens were irradiated in the BR2 reactor and located in the axial position of the reactor to reach four different neutron fluence levels with three tensile specimens at each position. All tests were performed at room temperature. The results clearly show that, within the composition ranges and experimental conditions considered herein, while copper has a dominant effect, phosphorus contribution is not significant. Moreover, no synergy between these two elements could be evidenced. For the second series of materials with variable nickel and manganese content, the latter does not seem to affect irradiation hardening for all compositions except the high nickel/high manganese steel. However, this last material exhibits an atypical behavior that can be attributed to deformation-induced martensitic transformation. This is supported by transmission electron microscopy examination of undeformed and deformed steels in the unirradiated and irradiated conditions and by experiments on pre-deformed and annealed tensile specimens.

[en] Exposure of austenitic stainless steels to liquid lead–bismuth eutectic with low concentration of dissolved oxygen typically results in selective leaching of highly-soluble alloying elements and ferritization of the dissolution-affected zone. In this work, focused ion beam, transmission electron backscatter diffraction and scanning transmission electron microscopy were utilized to elucidate early-stage aspects of the dissolution corrosion process of cold-worked austenitic stainless steels exposed to static, oxygen-poor liquid lead–bismuth eutectic at 450 °C for 1000 h. It was found that deformation-induced twin boundaries in the cold-worked steel bulk provide paths of accelerated penetration of the liquid metal into the steel bulk.

[en] Highlights: • Sigma phase was detected in solution annealed 316L by combining mechanical testing and in-depth microstructural analysis • Simulated diffraction patterns are used to validate low quality indexations of phases using EBSD • Additions to standard qualification procedures of steels containing high amounts of Cr and Mo are recommended MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is an accelerator driven system, currently under development at SCK CEN in Mol, Belgium. This nuclear system will use liquid Lead-Bismuth eutectic alloy as a spallation target for fast neutron production and as coolant. The ideal structural material for a liquid metal cooled reactor should be unsusceptible to both liquid metal embrittlement and liquid metal corrosion, while possessing high toughness. Nuclear grade austenitic stainless steels similar to AISI 316L have therefore been chosen as the main candidate structural materials for MYRRHA. In the framework of the qualification of those candidates, a specific batch of this material has shown unexpectedly poor mechanical properties, which triggered the need of in-depth microstructural analysis. The behaviour was attributed to the unexpected and undesired presence of intermetallic σ-phase. The σ-phase was identified with a high confidence by combining the data for chemical composition from energy-dispersive X-ray spectroscopy and crystallographic information from electron backscatter diffraction by comparing simulated Kikuchi diffraction patterns with experimentally recorded ones. At first sight the optical appearance of σ-phase resembles δ-ferrite islands, which results in the risk of overlooking this when only classical material qualification methods are used. When left undetected, testing this material including the brittle σ-phase in a liquid metal environment, in combination with miniature mechanical test specimens, could lead to misinterpretation of embrittlement of the austenitic matrix.

[en] Highlights: ► Gas atomization reaction synthesis used for precursor powders of ODS ferritic steel. ► Internal O exchange reactions were used to form Y-enriched nano-metric dispersoids. ► Final ODS microstructures develop from solidification features of initial powders. ► Particle size was shown as a viable method to control final ODS microstructure. ► Thermal–mechanical processing achieved desired fine-scale dislocation substructures. - Abstract: Gas atomization reaction synthesis was employed as a simplified method for processing oxide dispersion forming precursor Fe-based powders (e.g., Fe–Cr–Y–Hf). During this process a reactive atomization gas (i.e., Ar–O₂) was used to oxidize the powder surfaces during primary break-up and rapid solidification of the molten alloy. This resulted in envelopment of the powders by an ultra-thin (t < 50 nm) metastable Cr-enriched oxide shell that was used as a vehicle to transport oxygen into the consolidated microstructure. Subsequent elevated temperature heat treatment promoted thermodynamically driven oxygen exchange reactions between trapped films of Cr-enriched oxide and internal (Y, Hf)-enriched intermetallic precipitates, resulting in highly stable nano-metric mixed oxide dispersoids (i.e., Y–Hf–O) that were identified with X-ray diffraction. Transmission electron microscopy and atom probe tomography results also revealed that the size and distribution of the dispersoids were found to depend strongly on the original rapidly solidified microstructure. To exploit this, several oxide dispersion strengthened microstructures were engineered from different powder particle size ranges, illustrating microstructural control as a function of particle solidification rate. Additionally, preliminary thermal–mechanical processing was used to develop a fine scale dislocation substructure for ultimate strengthening of the alloy.