Compared to austenitic stainless steels, largely employed in the early fission reactors, high chromium Ferritic/Martenstic (FM) steels, developed since the first half of the 20th century for ...fossil-fuel power-plants, have a number of advantageous properties among which lower thermal expansion, higher thermal conductivity and better void swelling resistance. At the beginning of the 1970s, FM steels found their first nuclear application as wrapper and fuel cladding materials in sodium-cooled fast reactors. They are now the reference materials for in-vessel components of future fusion reactors, and are considered for in-pile and out-of-pile applications in Generation IV reactors as well as for various other nuclear systems. In this paper, after an introductory historical overview, the challenges associated with the use of FM steels in advanced reactors are addressed, including fabrication, joining and codification issues. The long term evolution of mechanical properties such as the creep and creep-fatigue behaviors is discussed and the degradation phenomena occurring in aggressive environments (lead alloys, high temperature gases, super-critical water and CO2, molten salts) are detailed. The paper also provides a brief overview of the radiation effects in FM steels. The influence of the key radiation parameters e.g. temperature, dose and dose rate on the microstructure and mechanical properties are discussed. The need to better understand the synergistic effects of displacement damage and helium produced by transmutation in fusion conditions is highlighted.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UL, UM, UPCLJ, UPUK
When the austenitic stainless steel 316L/N) was selected for ITER, it was well known that it would not be suitable for DEMO and fusion reactors due to its irradiation swelling at high doses. A ...parallel programme to ITER collaboration already had been put in place, under an IEA fusion materials implementing agreement for the development of a low activation ferritic/martensitic steel, known for their excellent high dose irradiation swelling resistance. After extensive screening tests on different compositions of Fe-Cr alloys, the chromium range was narrowed to 7-9% and the first RAFM was industrially produced in Japan (F82H: Fe-8%Cr-2%W-TaV). All IEA partners tested this steel and contributed to its maturity. In parallel several other RAFM steels were produced in other countries. From those experiences and also for improving neutron efficiency and corrosion resistance, European Union opted for a higher chromium lower tungsten grade, Fe-9%Cr-l%W-TaV steel (Eurofer), and in 1997 ordered the first industrial heats. Other industrial heats have been produced since and characterised in different states, including irradiated up to 80 dpa. China, India, Russia, Korea and US have also produced their grades of RAFM steels, contributing to overall maturity of these steels. This paper reviews the work done on RAFM steels by the fusion materials community over the past 30 years, in particular on the Eurofer steel and its design code qualification for RCC-MRx.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UL, UM, UPCLJ, UPUK
The spin transfer torque is essential for electrical magnetization switching. When a magnetic domain wall is driven by an electric current through an adiabatic spin torque, the theory predicts a ...threshold current even for a perfect wire without any extrinsic pinning. The experimental confirmation of this 'intrinsic pinning', however, has long been missing. Here, we give evidence that this intrinsic pinning determines the threshold, and thus that the adiabatic spin torque dominates the domain wall motion in a perpendicularly magnetized Co/Ni nanowire. The intrinsic nature manifests itself both in the field-independent threshold current and in the presence of its minimum on tuning the wire width. The demonstrated domain wall motion purely due to the adiabatic spin torque will serve to achieve robust operation and low energy consumption in spintronic devices.
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IJS, IZUM, KILJ, KISLJ, NUK, PILJ, PNG, SAZU, UL, UM, UPUK
Reduced activation ferritic/martensitic steels are currently the most technologically mature option for the structural material of proposed fusion energy reactors. Advanced next-generation higher ...performance steels offer the opportunity for improvements in fusion reactor operational lifetime and reliability, superior neutron radiation damage resistance, higher thermodynamic efficiency, and reduced construction costs. The two main strategies for developing improved steels for fusion energy applications are based on (1) an evolutionary pathway using computational thermodynamics modelling and modified thermomechanical treatments (TMT) to produce higher performance reduced activation ferritic/martensitic (RAFM) steels and (2) a higher risk, potentially higher payoff approach based on powder metallurgy techniques to produce very high strength oxide dispersion strengthened (ODS) steels capable of operation to very high temperatures and with potentially very high resistance to fusion neutron-induced property degradation. The current development status of these next-generation high performance steels is summarized, and research and development challenges for the successful development of these materials are outlined. Material properties including temperature-dependent uniaxial yield strengths, tensile elongations, high-temperature thermal creep, Charpy impact ductile to brittle transient temperature (DBTT) and fracture toughness behaviour, and neutron irradiation-induced low-temperature hardening and embrittlement and intermediate-temperature volumetric void swelling (including effects associated with fusion-relevant helium and hydrogen generation) are described for research heats of the new steels.
Reduced-activation ferritic-martensitic (RAFM) steels, candidate structural materials for fusion reactors, have achieved technological maturity after about three decades of research and development. ...The recent status of a few developmental aspects of current RAFM steels, such as aging resistance, plate thickness effects, fracture toughness, and fatigue, is updated in this paper, together with ongoing efforts to develop next-generation RAFM steels for superior high-temperature performance. In addition to thermomechanical treatments, including nonstandard heat treatment, alloy chemistry refinements and modifications have demonstrated some improvements in high-temperature performance. Castable nanostructured alloys (CNAs) were developed by significantly increasing the amount of nanoscale MX (M = V/Ta/Ti, X = C/N) precipitates and reducing coarse M23C6 (M = Cr). Preliminary results showed promising improvement in creep resistance and Charpy impact toughness. Limited low-dose neutron irradiation results for one of the CNAs and China low activation martensitic are presented and compared with data for F82H and Eurofer97 irradiated up to ∼70 displacements per atom at ∼300–325 °C.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UL, UM, UPCLJ, UPUK, ZRSKP
Several types of reduced activation ferritic/martensitic (RAFM) steel have been developed over the past 30years in China, Europe, India, Japan, Russia and the USA for application in ITER test blanket ...modules (TBMs) and future fusion DEMO and power reactors. The progress has been particularly important during the past few years with evaluation of mechanical properties of these steels before and after irradiation and in contact with different cooling media. This paper presents recent RAFM steel results obtained in ITER partner countries in relation to different TBM and DEMO options.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UL, UM, UPCLJ, UPUK
Based on the results of the IFMIF/EVEDA project in the broader approach (BA) activities, the conceptual design of the Advanced Fusion Neutron Source (A-FNS) to obtain irradiation data of DEMO blanket ...materials is under consideration in Japan. This paper describes the progress of the A-FNS conceptual design, which consists of a 40 MeV, 125 mA deuteron beam, a liquid lithium target system, and a test and post-irradiation test facilities. the main objectives of the A-FNS are the acquisition of RAFM irradiation data up to 2035 and the tritium recovery test on the blanket. In addition, it is expected to be used for various neutron researches. The technical design activities of the A-FNS are focused on improvement of the lithium target design, review of impurity criteria, remote handling, and irradiation application plan.
The status and key issues of reduced activation ferritic/martensitic (RAFM) steels R&D are reviewed as the primary candidate structural material for fusion energy demonstration reactor blankets. This ...includes manufacturing technology, the as-fabricated and irradiates material database and joining technologies. The review indicated that the manufacturing technology, joining technology and database accumulation including irradiation data are ready for initial design activity, and also identifies various issues that remain to be solved for engineering design activity and qualification of the material for international fusion material irradiation facility (IFMIF) irradiation experiments that will validate the data base.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UL, UM, UPCLJ, UPUK
The irradiation hardening behavior of reduced-activation ferritic steels after single Fe-ion beam irradiation and dual-ion (Fe ion and He ion) beam irradiation experiments was investigated with ...nanoindentation tests. The ion-irradiation experiments were conducted at 563K with 6.4MeV Fe3+ ions up to 3dpa at a 600nm depth from the irradiated surface. Furthermore, these experiments were conducted with and without simultaneous energy-degraded 1MeV He+ ions up to 300appm. The materials used were F82H, F82H+1Ni, and F82H+2Ni to investigate the effect of Ni addition on the irradiation hardening behavior. The measured nanoindentation hardness was converted to the bulk-equivalent hardness based on a combination of the Nix–Gao model to explain the indentation size effect and the composite hardness model to explain the softer substrate effect of the nonirradiated region beyond the irradiated depth range. It is clearly shown that the Ni addition enhances the irradiation hardening of F82H. The bulk-equivalent hardness is compared with the experimentally obtained Vickers hardness of F82H steels after neutron irradiation. The effect of simultaneously implanted helium on the irradiation hardening is negligible in the investigated irradiation conditions.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UL, UM, UPCLJ, UPUK
Understanding the effects of helium on microstructures and mechanical properties of reduced-activation ferritic-martensitic steels is important to use of these steels in fusion reactor structures. ...The 9Cr-2WVTa steels were doped with 58Ni and 60Ni isotopes at 2 weight percent to control the rate of transmutation helium generation. The samples were irradiated in the High Flux Isotope Reactor to ~24 displacements per atom at nominal temperatures of 300, 400, and 500°C, producing 228 and 7 atomic parts-per-million helium in the 58Ni- and 60Ni-doped samples, respectively. Transmission electron microscopy revealed a variety of precipitates and the radiation-induced dislocation loops and cavities (voids or helium bubbles). Tensile tests of the irradiated samples at the irradiation temperatures showed radiation-induced hardening at 300°C and radiation-induced softening at 400°C. Analysis indicates that the hardening primarily originated from the loops and cavities. The 58Ni-doped samples had greater strengthening contributions from loops and cavities, leading to higher hardening with lower ductility than the 60Ni-doped samples. The greater helium production of 58Ni did not show pronounced reductions in ductility of the samples.
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GEOZS, IJS, IMTLJ, KILJ, KISLJ, NUK, OILJ, PNG, SAZU, SBCE, SBJE, UL, UM, UPCLJ, UPUK
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