Sodium fast reactor evaluation: Core materials

In the framework of the Generation IV Sodium Fast Reactor (SFR) Program the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. In this paper the status of available and developmental materials for SFR core cladding and duct applications is reviewed. To satisfy the Generation IV SFR fuel requirements, an advanced cladding needs to be developed. The candidate cladding materials are austenitic steels, ferritic/martensitic (F/M) steels, and oxide dispersion strengthened (ODS) steels. A large amount of irradiation testing is required, and the compatibility of cladding with TRU-loaded fuel at high temperatures and high burnup must be investigated. The more promising F/M steels (compared to HT9) might be able to meet the dose requirements of over 200 dpa for ducts in the GEN-IV SFR systems.     

Response of nanoclusters in a 9Cr ODS steel to 1 dpa, 525 °C proton irradiation

Ferritic-martensitic (F/M) alloys are expected to play an important role as cladding or structural components in Generation IV and other advanced nuclear systems operating in the temperature range 350-700 °C and to doses up to 200 displacements per atom (dpa). Oxide dispersion strengthened (ODS) F/M steels have been developed to operate at higher temperatures than traditional F/M steels. These steels contain nanometer-sized Y-Ti-O nanoclusters for additional strengthening. A proton irradiation to 1 dpa at 525 °C has been performed on a 9Cr ODS steel to determine the nanocluster stability at low dose. The evolution of the nanocluster population and the composition at the nanocluster-matrix interface were studied using electron microscopy and atom probe tomography. The data from this study are contrasted to those from a previous study on heavy-ion irradiated 9Cr ODS steel.     

Corrosion-resistant coating technique for oxide-dispersion-strengthened ferritic/martensitic steel

Oxide-dispersion-strengthened (ODS) steels are attractive materials for application as fuel cladding in fast reactors and first-wall material of fusion blanket. Recent studies have focused more on high-chromium ferritic (12–18 wt% Cr) ODS steels with attractive corrosion resistance properties. However, they have poor material workability, require complicated heat treatments for recrystallization, and possess anisotropic microstructures and mechanical properties. On the other hand, low-chromium ferritic/martensitic (8–9 wt% Cr) ODS steels have no such limitations; nonetheless, they have poor corrosion resistance properties. In our work, we developed a corrosion-resistant coating technique for a low-chromium ferritic/martensitic ODS steel. The ODS steel was coated with the 304 or 430 stainless steel, which has better corrosion resistances than the low-chromium ferritic/martensitic ODS steels. The 304 or 430 stainless steel was coated by changing the canning material from mild steel to stainless steel in the conventional material processing procedure for ODS steels. Microstructural observations and micro-hardness tests proved that the stainless steels were successfully coated without causing a deterioration in the mechanical property of the low-chromium ferritic/martensitic ODS steel.     

先进反应堆用ODS F/M钢的强韧性匹配研究进展

氧化物弥散强化（ODS）铁素体/马氏体（F/M）钢具有极高的缺陷阱密度,加之体心立方结构基体,使其具有优异的抗辐照性能,被确定为包括聚变堆和第4代裂变堆在内的先进核能系统关键部件候选材料,成为核材料领域的研究热点。有利于ODS钢抗辐照性能的显微组织特点同时也赋予了ODS钢优异的室温和高温强度。但和其他高强度材料类似,ODS钢也存在强度高,而塑韧性不足的矛盾,不利于复杂部件的加工,因此,实现ODS F/M钢的高强高韧成为面向工程应用的ODS F/M钢研究的一大热点和难点。目前,针对ODS F/M钢强韧化的研究还较少,已有的相关研究也不够系统和深入,本文主要对抗辐照ODS F/M钢的显微组织结构要求及其强韧化研究现状进行总结和分析,为先进反应堆用抗辐照ODS F/M钢的强韧化设计提供思路和参考。     

Radiation response of a 9 chromium oxide dispersion strengthened steel to heavy ion irradiation

Ferritic-martensitic (FM) alloys are expected to play an important role as cladding or structural components in Generation IV systems operating in the temperature range 350-700 °C and to doses up to 200 dpa. Oxide dispersion strengthened (ODS) ferritic-martensitic steels have been developed to operate at higher temperatures than traditional FM steels. These steels contain nanometer-sized Y-Ti-O nanoclusters as a strengthening mechanism. Heavy ion irradiation has been used to determine the nanocluster stability over a temperature range of 500-700 °C to doses of 150 dpa. At all temperatures, the average nanocluster size decreases but the nanocluster density increases. The increased density of smaller nanoclusters under radiation should lead to strengthening of the matrix. While a reduction in size under irradiation has been reported in some other studies, many report oxide stability. The data from this study are contrasted to the available literature to highlight the differences in the reported radiation response.     

Defect and void evolution in oxide dispersion strengthened ferritic steels under 3.2 MeV Fe ion irradiation with simultaneous helium injection

In an attempt to explore the potential of oxide dispersion strengthened (ODS) ferritic steels for fission and fusion structural materials applications, a set of ODS steels with varying oxide particle dispersion were irradiated at 650°C, using 3.2 MeV Fe⁺ and 330 keV He⁺ ions simultaneously. The void formation mechanisms in these ODS steels were studied by juxtaposing the response of a 9Cr–2WVTa ferritic/martensitic steel and solution annealed AISI 316LN austenitic stainless steel under the same irradiation conditions. The results showed that void formation was suppressed progressively by introducing and retaining a higher dislocation density and finer precipitate particles. Theoretical analyses suggest that the delayed onset of void formation in ODS steels stems from the enhanced point defect recombination in the high density dislocation microstructure, lower dislocation bias due to oxide particle pinning, and a very fine dispersion of helium bubbles caused by trapping helium atoms at the particle–matrix interfaces.     

Bootstrap calculation of ultimate strength temperature maxima for neutron irradiated ferritic/martensitic steels

The dependence of mechanical properties of ferritic/martensitic (F/M) steels on irradiation temperature is of interest because these steels are used as structural materials for fast, fusion reactors and accelerator driven systems. Experimental data demonstrating temperature peaks in physical and mechanical properties of neutron irradiated pure iron, nickel, vanadium, and austenitic stainless steels are available in the literature. A lack of such an information for F/M steels forces one to apply a computational mathematical-statistical modeling methods. The bootstrap procedure is one of such methods that allows us to obtain the necessary statistical characteristics using only a sample of limited size. In the present work this procedure is used for modeling the frequency distribution histograms of ultimate strength temperature peaks in pure iron and Russian F/M steels EP-450 and EP-823. Results of fitting the sums of Lorentz or Gauss functions to the calculated distributions are presented. It is concluded that there are two temperature (at 360 and 390 °C) peaks of the ultimate strength in EP-450 steel and single peak at 390 °C in EP-823.     

Analysis of hardening limits of oxide dispersion strengthened steel

Oxide dispersion strengthened (ODS) ferritic/martensitic (F/M) steels are promising materials for high temperature applications. The hardening limits from room temperature to 1000 °C of one of such steel, ODS EUROFER97, together with the impact of the material production steps, are investigated at a microstructural level by coupling hardness, tensile tests and transmission microscopy, including in situ heating experiments. The oxides, ytttria and complex yttrium titanium oxides, reinforce the material by forming more or less stable obstacles to dislocations, and by promoting grain refinement by pinning grain boundaries. It appears that part of the yttrium titanium oxides particles dissolves from about 600 °C while pure yttria particles are stable at least to 1000 °C in the steel. The concurrent roles of the oxides and the dislocation structure in the hardening are rationalized using the dispersion barrier hardening model. It appears that hardening due to dislocations can overcome the one due to oxides but is more sensitive to temperature than the one due to oxides, and that the main limiting factor is the thermal stability of the oxides.     

先进快堆外套管材料铁素体/马氏体钢研究进展

介绍了快堆外套管材料的使用要求以及国内外快堆外套管材料的选用情况,论述了快堆中中子辐照对铁素体/马氏体钢的微观结构和力学性能的影响,并介绍了国产快堆外套管材料铁索体/马氏体钢的研发计划.铁索体/马氏体钢具有优异的抗辐照肿胀性能,T92钢作为第三代铁索体/马氏体钢同时具有良好的高温强度,被作为国产快堆外套管的首要候选材料...     

Behavior of ferritic/martensitic steels after n-irradiation at 200 and 300 °C

High chromium ferritic/martensitic (F/M) steels are considered as the most promising structural materials for accelerator driven systems (ADS). One drawback that needs to be quantified is the significant hardening and embrittlement caused by neutron irradiation at low temperatures with production of spallation elements. In this paper irradiation effects on the mechanical properties of F/M steels have been studied and comparisons are provided between two ferritic/martensitic steels, namely T91 and EUROFER97. Both materials have been irradiated in the BR2 reactor of SCK-CEN/Mol at 300 °C up to doses ranging from 0.06 to 1.5 dpa. Tensile tests results obtained between −160 °C and 300 °C clearly show irradiation hardening (increase of yield and ultimate tensile strengths), as well as reduction of uniform and total elongation. Irradiation effects for EUROFER97 starting from 0.6 dpa are more pronounced compared to T91, showing a significant decrease in work hardening. The results are compared to our latest data that were obtained within a previous program (SPIRE), where T91 had also been irradiated in BR2 at 200 °C (up to 2.6 dpa), and tested between −170 °C and 300 °C. Irradiation effects at lower irradiation temperatures are more significant.     

Development of Oxide Dispersion Strengthened Steels for FBR Core Application, (II)

Previously manufactured oxide dispersion strengthened (ODS) ferritic steel cladding tubes had inferior internal creep rupture strength in the circumferential hoop direction. This unexpected feature of ODS cladding tubes was substantially ascribed to the needle-like grain structure aligned with the forming direction. In this study, the grain morphology was controlled by using the martensite transformation in ODS martensitic steels to produce an equi-axial grain structure. A major improvement in the strength anisotropy was successfully achieved. The most effective yttria addition was about 1 mass% in improving the strength of the ODS martensitic steels. A simple addition of titanium was particularly effective in increasing the strength level of the ODS martensitic steels to that of ODS ferritic steels.     

Interfacial microstructures and hardness distributions of vacuum plasma spraying W-coated ODS ferritic steels for fusion plasma facing applications

In the present study, interfacial microstructures and hardness distributions of W-coated ODS steels as plasma facing structural materials were investigated. A vacuum plasma spraying (VPS) technique was employed to fabricate a W layer on the surface of the ODS ferritic steel substrates. The microstructural observations revealed that the VPS-W has very fine grains aligned toward the spraying direction, and a favorable interface between W and ODS ferritic steels by a mechanical inter-locking without an intermetallic layer. However, crack-type defects were found in VPS-W. Because a brittle inter-diffused layer does not exist at the joint interface, the hardness was gradually distributed in the joint region. After neutron irradiation, irradiation hardening significantly occurred in the VPS-W. However, the hardening of VPS-W was less than that of bulk W irradiated at 773 K. Thus, the VPS is considered to be one of the promising ways to join dissimilar materials between W and ODS steels, which can avoid the formation of an interfacial intermetallic layer and create favorable irradiation hardening resistance on the W coated layer.     

Microstructural evolution of Y2O3 and MgAl2O4 ODS EUROFER steels during their elaboration by mechanical milling and hot isostatic pressing

Different ODS EUROFER steels reinforced with Y₂O₃ and MgAl₂O₄ were elaborated by mechanical milling and hot isostatic pressing. Good compromise between strength and ductility could be obtained but the impact properties remain low (especially for the Y₂O₃ ODS steel). The materials were structurally characterized at each step of the elaboration. During milling, the martensite laths of the steel are transformed into nano-metric ferritic grains and the Y₂O₃ oxides dissolve (but not the MgAl₂O₄ spinels). After the HIP, all the ODS steels remain ferritic with micrometric grains, surrounded by nano-metric grains for the Y₂O₃ ODS steels. The mechanisms in the Y₂O₃ ODS steels are complex: the Y₂O₃ oxides re-precipitate as nano-Y₂O₃ particles that impede a complete austenitization during the HIP. The quenchability of the ODS steels is modified by the milling process, the oxide nature and the oxide content. Eventually, the advantages and drawbacks of each oxide type are discussed.     

High heat flux testing of 12-14Cr ODS ferritic steels

The thermal performance of Fe-(12-14)Cr-2W-0.3Ti-0.3Y₂O₃ ODS reduced activation ferritic steels, which are considered as candidate first wall materials for the future fusion power reactors and were manufactured by mechanical alloying in hydrogen and hot isostatic pressing, was assessed by high heat flux (HHF) testing with the electron beam JUDITH facility at the Forschungszentrum Jülich (FZJ), Germany. An analysis of the microhardness and microstructure of the specimens was done before and after HHF tests.In general, both materials present a ferritic (α-Fe, bcc) microstructure with a wide range of grain sizes from 100 to 500 nm up to a few micrometers. The coarse grains are almost dislocation-free, while the smaller ones are surrounded by tangles of dislocations. Oxide and carbide impurities (about a few hundreds nm in size) and a high density of Y-Ti-O nano-clusters, with a mean size of about 5 nm, are also present. The microhardness, density and tensile strength of the 14Cr material are slightly larger than those of the 12Cr material.HHF tests revealed that there is no difference in thermal performance, level of degradation and erosion behaviour of 12Cr and 14Cr ODS steels. The onset of melting of the materials occurs for an energy density between 1 and 1.5 MJ/m². Below this value only some kind of thermal etching takes place. This is a significant improvement compared to stainless steel, for which severe plastic deformation at the material surface was observed.     

Status of materials handbooks for particle accelerator and nuclear reactor applications

In support of research and development for accelerator applications, a materials handbook was developed in August of 1998 funded by the Accelerator Production of Tritium Project. This handbook, presently called Advanced Fuel Cycle Initiative (AFCI) Materials Handbook, Materials Data for Particle Accelerator Applications, has just issued Revision 5 and contains detailed information showing the effects of irradiation on many properties for a wide variety of materials. Development of a web-accessible materials database for Generation IV Reactor Programs has been ongoing for about three years. This handbook provides a single authoritative source for qualified materials data applicable to all Generation IV reactor concepts. A beta version of this Gen IV Materials Handbook has been completed and is presently under evaluation.     

Nano-indentation measurement of oxide layers formed in LBE on F/M steels

Ferritic/martensitic (F/M) steels (T91, HT-9, EP 823) are candidate materials for future liquid lead or lead bismuth eutectic (LBE) cooled nuclear reactors. To understand the corrosion of these materials in LBE, samples of each material were exposed at 535 °C for 600 h and 200 h at an oxygen content of 10⁻⁶ wt%. After the corrosion tests, the samples were analyzed using SEM, WDX and nano-indentation in cross section. Multi-layered oxide scales were found on the sample surfaces. The compositions of these oxide layers are not entirely in agreement with the literature. The nano-indentation results showed that the E-modulus and hardness of the oxide layers are significantly lower than the values for dense bulk oxide materials. It is assumed that the low values stem from high porosity in the oxide layers. Comparison with in-air oxidized steels show that the E-modulus decreases with increasing oxide layer thickness.     

Corrosion experiments of steels in flowing Pb at 500 °C and in flowing LBE at 450 °C

In accelerator driven systems (ADS), as well as in the next Generation IV reactors, one of the concerned issues is the material compatibility and corrosion in liquid Pb, which is considered a candidate coolant. Liquid metal corrosion of the structural materials can proceed via different processes: species dissolution and penetration of liquid metal along grain boundaries and metal. The occurrence of these corrosion phenomenon depend on the experimental parameters, such as temperature, thermal gradients, solid and liquid metal compositions, velocity of the liquid metal and oxygen activity in Pb. One possible technique to prevent any corrosive attack by the liquid metals is the in situ passivation of the containment steels. This technique is achieved through an active control and monitoring of the dissolved oxygen concentration. This paper summarizes the data gathered from the CHEOPE III loop, where passivation of T91 and AISI 316L steels is tested in pure Pb at 500 °C were carried out, comparing them with preliminary corrosion data, in LBE, gathered from the LECOR loop.     

Si含量对9%Cr铁素体马氏体钢微观结构和力学性能的影响

铁素体马氏体（F/M）钢是铅冷快堆堆芯的主要候选材料之一,提高Si含量可提高其抗腐蚀性能,但影响其微观结构和力学性能。为研发兼顾力学性能和抗腐蚀性能的高Si含量F/M钢,本文对Si含量（质量分数为034%、061%、080%、098%和120%）对9%Cr F/M钢微观结构和力学性能的影响进行了研究。结果显示,室温下通过正火（1 050 ℃,05 h,空冷）+回火（760 ℃,15 h,空冷）热处理制备的F/M钢均为全马氏体组织。钢的屈服强度和抗拉强度均随Si含量的增加而增加。Si含量对钢的组织及冲击性能的影响可分为两个阶段,当F/M钢中Si含量为034%~080%时,其组织、冲击性能变化很小;Si含量为098%~120%时,F/M钢中的第二相尺寸、数量增加,沿马氏体板条界面析出少量的Laves相,F/M钢的冲击韧性降低。本研究9%Cr F/M钢中Si元素的最优添加量应低于08%,使钢在保持较高强度的同时兼具良好的韧性。     

European cross-cutting research on structural materials for Generation IV and transmutation systems

It has been internationally recognized that materials science and materials development are key issues for the implementation of innovative reactor systems such as those defined in the framework of the Generation IV and advanced fuel cycle initiatives. In Europe, materials studies are considered within the Strategic Research Agenda of the Sustainable Nuclear Energy Technology Platform. Moreover, the European Commission has recently launched a 7th Framework Programme Research Project, named ‘Generation IV and Transmutation Materials’, that has the objective of addressing materials issues which are cross-cutting for more than one type of innovative reactor systems. The present work has been prepared with the aim of describing the rationale, the objectives, the work plan and the expected results of this research project.     

Characterization of recrystallization of 12Cr and 15Cr ODS ferritic steels

The recrystallization behavior of 12Cr and 15Cr oxide dispersion-strengthened (ODS) ferritic steels, which are the promising candidate materials for long-life core materials of the advanced fast breeder reactors, was investigated in terms of an intermediate softening heat treatment. It was clarified that keeping recovery structure at the intermediate heat treatment is indispensable for producing recrystallized structure at the final heat treatment. Prevention of repeating recrystallization is owing to the stable {100} 〈110〉 texture formation with less stored strain energy by the cold-rolling of the recrystallized structure. The two-step softening process was proposed to suppress the recrystallization and obtain adequate hardness reduction at the intermediate heat treatment. This process is effective for producing a stable recrystallized structure at the final heat treatment of the manufacturing process of ODS ferritic steel cladding.