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.     

硅对9Cr-1.5WVTa低活化马氏体钢力学性能的影响

基于中国正在研究的聚变堆用9Cr-1.5WVTa低活化马氏体钢(CLAM钢),研究了添加合金元素硅对CLAM钢力学性能的影响。结果表明,添加0.2%Si使得CLAM钢的抗拉强度和屈服强度明显提高,钢的塑性和冲击韧性同时也得到一定提高,其中,韧脆转变温度(DBTT)由-13℃降至-30℃。未添加和添加0.29%Si的CLAM钢均为全马氏体组织,无δ铁素体存在。硅的添加使得9Cr-1.5WVTaSi钢的晶粒细化,从而提高了钢的拉伸和冲击性能。     

硅对低活化马氏体钢电子辐照行为的影响

利用超高压透射电子显微镜研究了两种成分的低活化马氏体钢(CLAM钢)的辐照损伤行为。结果表明：电子辐照能在未添加硅的CLAM钢中产生辐照空洞;在450℃下辐照至14dpa时,空洞数密度约为8.7×10²¹m^－3,辐照肿胀率约为0.26％;在450℃下的辐照肿胀率明显比500℃下的高;当损伤率为2×10^－3dpa/s时,添加合金元素硅能显著提高CLAM钢的抗辐照肿胀能力,未在添加硅的CLAM钢中实验观察到辐照空洞的形成。在450℃下进行辐照时,添加硅的CLAM钢出现明显的辐照共格析出现象。     

9Cr3W低活性马氏体钢微观组织及力学性能的初步研究

通过提高W含量,调整V、Ta、Ti、N等微合金元素含量,设计了9Cr3W型低活性马氏体钢。研究了该钢的微观组织结构与硬化、时效及相转变行为,对其进行了拉伸性能和冲击韧性测试。与Eurofer97钢相比,该钢表现出优良的高温拉伸性能。分析了9Cr3W钢用作超临界水堆堆芯内部件及包壳材料的可行性,其高温力学性能远优于Zr合金包壳材料;拉伸性能与T91钢相当,且韧脆转变温度低于T91钢,冲击吸收功上限高于T91钢,具有优良的冲击韧性;9Cr3W钢的高温瞬时强度低于奥氏体316不锈钢,成为制约其用于超临界水堆堆芯内部件及包壳的因素之一。     

Corrosion of stainless steels in lead-bismuth eutectic up to 600 °C

An experimental program has been carried out to understand the differences in the corrosion behaviour between different stainless steels: the austenitic steels 304L and 316L, the martensitic steels F82Hmod, T91 and EM10, and the low alloy steel P22. The influence of oxygen level in Pb-Bi, temperature and exposure time is studied. At 600 °C, the martensitic steels and the P22 steel exhibit thick oxide scales that grow with time, following a linear law for the wet environment and a parabolic law for the dry one. The austenitic stainless steels show a better corrosion behaviour, especially AISI 304L. Under reducing conditions, the steels exhibit dissolution, more severe for the austenitic stainless steels. At 450 °C, all the materials show an acceptable behaviour provided a sufficient oxygen level in the Pb-Bi. At reducing conditions, the martensitic steels and the P22 steel have a good corrosion resistance, while the austenitic steels exhibit already dissolution at the longer exposures.     

Analysis of cleavage fracture potential of martensitic stainless steel fusion structures: Part I. Micromechanical models and material properties

The major disadvantage of martensitic stainless steels for structural applications in fusion reactors is currently considered to be their potential for low temperature brittle cleavage fracture. This study attempts to review the current understanding of cleavage fracture in steels and the role of microstructure in dictating material resistance to this type of fracture. A parametric analysis of cleavage fracture in a surrogate steel, A533B, is made and the results are used in conjunction with general cleavage fracture theory to establish some potential guidelines for future research in developing the martensitic stainless steels.     

Analysis of cleavage fracture potential of martensitic stainless steel fusion structures: Part II. Fracture analysis procedures for flawed fusion structures

The potential for brittle cleavage fracture is a major concern for martensitic stainless steels which are candidates for fusion reactor structural materials. This study attempts to identify for flawed fusion structures the pertinent fracture resistance or failure parameters and the relationships between these parameters and the basic materials properties which govern cleavage fracture. Several procedures for relating test data to failure prediction, including Charpy-V-notch transition temperature referencing and two-parameter interpolation procedures, are considered; and results are discussed with respect to possible research paths for martensitic stainless steel alloy development.     

Temperature effects on the mechanical properties of candidate SNS target container materials after proton and neutron irradiation

This report presents the tensile properties of EC316LN austenitic stainless steel and 9Cr-2WVTa ferritic/martensitic steel after 800 MeV proton and spallation neutron irradiation to doses in the range 0.54-2.53 dpa at 30-100 °C. Tensile testing was performed at room temperature (20 °C) and 164 °C. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr-2WVTa ferritic/martensitic steel posted negative hardening in the engineering stress-strain curves. In the EC316LN stainless steel, increasing the test temperature from 20 to 164 °C decreased the strength by 13-18% and the ductility by 8-36%. The effect of test temperature for the 9Cr-2WVTa ferritic/martensitic steel was less significant than for the EC316LN stainless steel. In addition, strain-hardening behaviors were analyzed for EC316LN and 316L stainless steels. The strain-hardening rate of the 316 stainless steels was largely dependent on test temperature. A calculation using reduction of area measurements and stress-strain data predicted positive strain hardening during plastic instability.     

An exploratory study to determine applicability of nano-hardness and micro-compression measurements for yield stress estimation

The superior properties of ferritic/martensitic steels in a radiation environment (low swelling, low activation under irradiation and good corrosion resistance) make them good candidates for structural parts in future reactors and spallation sources. While it cannot substitute for true reactor experiments, irradiation by charged particles from accelerators can reduce the number of reactor experiments and support fundamental research for a better understanding of radiation effects in materials. Based on the nature of low energy accelerator experiments, only a small volume of material can be uniformly irradiated. Micro and nanoscale post irradiation tests thus have to be performed. We show here that nanoindentation and micro-compression testing on T91 and HT-9 stainless steel before and after ion irradiation are useful methods to evaluate the radiation induced hardening.     

State of the development and application techniques of the steel X 20 CrMoV 12 1

In most European and many other countries, the martensitic steel X 20 CrMoV 12 1 is used for the most highly stressed components of conventional power stations, such as superheater tubing and main steam pipe systems. This steel permits the use of much thinner wall thicknesses and a more flexible design as compared with lower alloyed steels. In the operation of such plants, there is the additional advantage of higher start-up velocities as opposed to the heavy wall piping systems. Accordingly, within the variety of high temperature steels, the X 20 CrMoV 12 1 steel is found to be the optimum link between the low-alloyed ferritic-bainitic and the high-alloyed austenitic steels.From the viewpoint of alloying, the high creep rupture strength of this steel is only attained by 12% chromium and the addition of molybdenum and vanadium. The high hardness level of the martensite formed in this steel following hot forming or welding has to be taken into account in further processing.     

Xe~+离子辐照诱导国产T91及316Ti钢微结构变化研究

对国产T91及316Ti钢进行室温下200keV的Xe~+离子辐照,使用X射线衍射(XRD)、透射电镜(TEM)等检测方法研究不同损伤剂量下辐照对材料相的稳定性和微观结构变化的影响。研究结果表明:T91钢辐照后未发生明显相变,而316Ti发生了γ(FCC)→α(BCC)的马氏体相变,且随辐照损伤剂量的增加,α相含量增加,相变的主要驱动力为辐照离子在辐照层的聚集从而产生的剪切应力;T91钢中的M_(23)C_6颗粒随辐照损伤剂量的增加,非晶化越来越明显,主要是由于辐照粒子的轰击削弱了M_(23)C_6颗粒晶格的稳定性,晶格塌陷成为非晶状态;316Ti钢在较低辐照损伤剂量(4.6dpa)下出现黑斑结构,而在高辐照损伤剂量(37.1dpa)下黑斑结构进一步聚集形成位错环。     

Effect of contact conditions on embrittlement of T91 steel by lead-bismuth

The T91 martensitic steel is a candidate structural material for the liquid lead-bismuth eutectic (LBE) MEGAPIE spallation target. This paper first reviews some results on Liquid Metal Embrittlement (LME) of martensitic steels by liquid metals. It appears that LME of steels can occur provided a few criteria are fulfilled simultaneously. Intimate contact between liquid metal and solid metal is the first one. Usually, it is impossible to avoid the oxide film formation on the steel surface even after short exposure to air. This explains the difficulty arising when one would like to determine the susceptibility to LME of T91 steel whilst put into contact with lead-bismuth. Later, we report on different methods of surface preparation in order to remove the oxide layer on the T91 steel (PVD, soft soldering fluxes) and the resulting susceptibility to LME.     

Sodium effects on mechanical performance and consideration in high temperature structural design for advanced reactors

Sodium environmental effects are key limiting factors in the high temperature structural design of advanced sodium-cooled reactors. A guideline is needed to incorporate environmental effects in the ASME design rules to improve the performance reliability over long operating times. This paper summarizes the influence of sodium exposure on mechanical performance of selected austenitic stainless and ferritic/martensitic steels. Focus is on Type 316SS and mod.9Cr-1Mo. The sodium effects were evaluated by comparing the mechanical properties data in air and sodium. Carburization and decarburization were found to be the key factors that determine the tensile and creep properties of the steels. A beneficial effect of sodium exposure on fatigue life was observed under fully reversed cyclic loading in both austenitic stainless steels and ferritic/martensitic steels. However, when hold time was applied during cyclic loading, the fatigue life was significantly reduced. Based on the mechanical performance of the steels in sodium, consideration of sodium effects in high temperature structural design of advanced fast reactors is discussed.     

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.     

Hydrogen transport and solubility in 316L and 1.4914 steels for fusion reactor applications

Equations are given which describe the permeation rate, diffusivity and solubility of hydrogen over the range 250–600°C at pressures up to 10⁵Pa for the 316L stainless and modified 1.4914 martensitic candidate steels proposed for the construction of the Next European Torus (NET). For heat-treated 316L steel, the permeation rates measured agreed well with previous work and did not vary significantly from specimen to specimen or from batch to batch.

Measurements of the permeation rate of hydrogen and deuterium through the modified 1.4914 steel, believed to be the first made, show that the martensitic steel is significantly more permeable than the austenitic steel, by an order of magnitude at 250°C and a factor of five at 600°C. This difference could make it necessary to use permeation barriers on critical components made from the martensitic steel in order to reduce the tritium permeation rate to acceptable levels.     

On the (in)adequacy of the Charpy impact test to monitor irradiation effects of ferritic/martensitic steels

Irradiation embrittlement studies rely very often on Charpy impact data, in particular the ductile-to-brittle transition temperature (DBTT). However, while the DBTT-shift is equivalent to the increase of the fracture toughness transition temperature of ferritic steels, it is not the case for ferritic/martensitic steels. The aim of this study is to critically assess experimental data obtained on a 9%Cr-ferritic/martensitic steel, Eurofer-97, to better understand the underlying mechanisms involved during the fracture process. More specifically, a dedicated analysis using the load diagram approach allows to unambiguously reveal the actual effects of irradiation on physically rather than empirically based parameters. A comparison is made between a ferritic and ferritic/martensitic steel to better identify the possible similarities and differences. Tensile, Charpy impact and fracture toughness tests data are examined in a global approach to assess the actual rather than apparent irradiation effects. The adequacy or inadequacy of the Charpy impact test to monitor irradiation effects is extensively discussed.     

低活化马氏体钢的微观结构与力学性能

介绍了作为聚变反应堆候选结构材料的低活化马氏体钢的基本设计思路,初步确定了材料的化学成分和热处理工艺,研究了材料的冶金特性、微观组织和力学性能.同时,对比了添加少量钇和硅对材料性能的影响,发现添加硅可以提高材料强度,同时能保证材料具有足够的塑性和韧性;钇的添加对改善材料的塑性很有帮助,但是会使材料强度降低.     

Mechanical properties of martensitic steels after exposure to flowing liquid metals

This paper describes the tensile test results of martensitic steels EUROFER 97 and T91 after exposure to flowing Pb-17Li and Pb-55.5Bi alloy, respectively. The corrosion tests were performed in LIFUS II loop (Pb-17Li) and LECOR loop (Pb-55.5Bi) located in the ENEA Brasimone Center. Tensile test were carried in SYNTECH test machine under Ar atmosphere with an extension rate of 2 mm/min, and test temperature for EUROFER and T91 steels were at 480 and 400 °C respectively. The exposure of EUROFER steel to the liquid Pb-17Li did not affect its mechanical properties, while the ductility of T91 steel was deteriorated after exposed to flowing liquid Pb-55.5Bi.     

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.     

Vacuum plasma-sprayed tungsten on EUROFER and 316L: Results of characterisation and thermal loading tests

Vacuum plasma-spraying (VPS) can be used for the industrial deposition of thick W coatings on actively water-cooled components made of low activation steel or stainless steel. Mock-ups made of martensitic steels, EUROFER and F82H, as well as steel 316L, were coated with 2 mm thick W-VPS layers. The coated materials are candidates for first wall components (ITER and DEMO) receiving moderate heat load of up to 1 MW/m². Mixed tungsten/steel interlayers were applied to reduce the residual and thermal stresses at the substrate–coating interface and to improve the adhesion of the coating. The advantage of this mixed interlayer is that no further (high activation) materials have to be introduced to improve coating adhesion.The characterisation of the W-VPS layers includes the evaluation of the coating microstructure, the measurement of physical and mechanical properties and the metallographical examination before and after heat load tests.Heat load tests with steady state operation up to 2.5 MW/m² and cycling heat loads of 2 MW/m², were successfully completed. They confirm the thermomechanical suitability of industrially manufactured W-VPS coatings for plasma facing first wall components made of steel.