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.     

Reduced-activation steels: Future development for improved creep strength

Reduced-activation steels for fusion applications were developed in the 1980s to replace the elevated-temperature commercial steels first considered. The new steels were patterned after the commercial steels, with the objective that the new steels have yield stress and ultimate tensile strength and impact toughness in a Charpy test comparable to or better than the steels they replaced. That objective was achieved in reduced-activation steels developed in Japan, Europe, and the United States. Although tensile and impact toughness of the reduced-activation steels exceed those of the commercial steels they were patterned after, their creep-rupture properties are inferior to some commercial steels they replaced. They are even more inferior to commercial steels developed since the 1980s. In this paper, compositional differences between reduced-activation steels and new commercial steels are examined, and compositions are proposed for development of new-and-improved reduced-activation steels.     

Chromium-molybdenum steels for fusion reactor first walls — a review

Chromium-molybdenum steels have recently become of interest as a first wall and blanket structural material for fusion reactors. This application will be assessed, and possible approaches on how Cr---Mo alloys may be further developed for this application will be proposed.Generally, the Cr---Mo steels can be divided into two categories: unmodified, basically Cr---Mo---C steels and Cr---Mo---C steels modified by the addition of carbide-forming elements. Extensive research and development efforts have been conducted on the unmodified steels, especially Cr-1 Mo and 12 Cr---Mo steels. Considerable work has also been done on 12 Cr---Mo steels modified with additions of V, Nb, Ti and W. In recent years much of the research effort on this type of alloy has been directed at developing modified Cr---Mo steels with less than 12% Cr (generally 9%) for applications where the “stainless” properties imparted by chromium additions of at least 12% are not needed.We will examine the unmodified and modified steels in terms of hardenability, precipitation processes (stability at elevated temperatures), strength, and toughness. Where possible, we will discuss the effects of irradiation on these properties. Such a study leads to the types of tradeoffs that may be necessary between the well-researched unmodified Cr-1 Mo steel and a high-chromium modified steel.     

Underwater explosive welding of tungsten to reduced-activation ferritic steel F82H

The present study reports the underwater explosive welding of commercially pure tungsten onto the surface of a reduced-activation ferritic steel F82H plate. Cross-sectional observation revealed the formation of a wave-like interface, consisting of a thin mixed layer of W and F82H. The results of nanoindentation hardness testing identified a gradual progressive change in the interface, with no hardened or brittle layer being observed. Small punch tests on the welded specimens resulted in cracking at the center of the tungsten, followed by crack propagation toward both the tungsten surface and the tungsten/steel interface.     

Effect of specimen size on the ductile-brittle transition behavior and the fracture sequence of 9Cr-W steels

The effect of specimen size on the ductile-brittle transition behavior and the fracture sequence were investigated by means of Charpy absorbed energy measurement and fractography, using the full size, the half size and the one-third size V-notch specimens of 9Cr-W steels. The steels used are reduced-activation ferritic steels for fusion reactor structures. Attempts were made to correlate the impact data between the different specimen sizes by using normalizing parameters, such as nominal fracture area and nominal fracture volume for the upper shelf energy and ligament size for the ductile-brittle transition temperature. Fractography showed a similar fracture sequence for the three different sizes of the specimens.     

超临界水冷堆燃料包壳管用低活性F/M钢的优化设计

应用热力学计算与实验验证,系统研究了Cr、W、C、Mn对高Cr低活性F/M（铁素体/马氏体）钢基体相及显微组织的影响规律,并在此基础上,对钢的组织和成分进行设计与优化,以适应超临界水系统对包壳材料的性能要求。研究表明：Cr是决定高Cr低活性实验钢中奥氏体Cr固溶量以及钢中是否出现铁素体的最重要影响因素;W和C对实验钢铁素体相的出现有显著影响,而Mn的影响相对较小;W对实验钢中Laves相出现的温度范围及数量具有显著影响,Laves相消失的临界温度随W量降低而降低;在不采用Co、Ni等奥氏体形成元素且不增加Mn量的情况下,通过调控W、C等含量,Cr含量≥11%的Cr-W-C-Mn系低活性F/M钢即可获得全马氏体组织。     

Composition influences and interactions in radiation sensitivity of reactor vessel steels

A series of 4-way split laboratory melts of low alloy, pressure vessel steel, representing statistical combinations of specific impurity elements and/or alloying elements, are being evaluated after 288°C irradiation to 2 × 10¹⁹ n/cm², E > 1 MeV. The objective is to reveal interactions between elements (or interdependencies) in radiation sensitivity development. The primary base composition for melting was ASTM A 302-B or A 533-B steel; plates from the melt casts were heat treated to simulate 150 mm or thicker plate materials.This report summarizes preirradiation-postirradiation notch ductility determinations for six metls (24 composition variations) produced for the study. The melts were used to test the effects on radiation sensitivity of phosphorus impurities in combination with copper impurities and the significance of high/low levels of nickel, manganese, molybdenum and chromium alloying to steels containing a significant copper content ( 0.15 % Cu). Contributions of tin and arsenic impurities were also evaluated.One key finding is an inverse dependence of the phosphorus contribution to radiation sensitivity level, on copper content. Therefore, for the new (improved) steels containing a low copper content, phosphorus level should not be ignored in making projections of radiation sensitivity in service. Other important interactions between elements are reported.     

Metallography studies and hardness measurements on ferritic/martensitic steels irradiated in STIP

In this work metallography investigations and microhardness measurements have been performed on 15 ferritic/martensitic (FM) steels and 6 weld metals irradiated in the SINQ Target Irradiation Program (STIP). The results demonstrate that all the steels have quite similar martensite lath structures. However, the sizes of the prior austenite grain (PAG) of these steels are quite different and vary from 10 to 86 μm. The microstructure in the fusion zones (FZ) of electron-beam welds (EBWs) of 5 steels (T91, EM10, MANET-II, F82H and Optifer-IX) is similar in respect to the martensite lath structure and PAG size. The FZ of the inert-gas-tungsten weld (TIGW) of the T91 steel shows a duplex structure of large ferrite gains and martensite laths. The microhardness measurements indicate that the normalized and tempered FM steels have rather close hardness values. The unusual high hardness values of the EBW and TIGW of the T91 steel were detected, which suggests that these materials are without proper tempering or post-welding heat treatment.     

Recent applications of small-angle neutron scattering in the characterization of irradiated steels for nuclear technologies

Small-angle neutron scattering (SANS) is a powerful experimental tool to investigate the microstructural evolution under irradiation in steels for fission and future fusion reactor systems. We present recent SANS results concerning the modelling of helium bubble growth in F82H-mod. steel implanted with α-particles and the dose dependence of microstructural radiation damage in Eurofer-97 steel for fusion reactors irradiated at 250 °C. The discussion of these results is focussed on the quality of the metallurgical information obtained by such SANS measurements and consequently on their usefulness also for engineering and design purposes.     

Method for Calculating and Optimizing a Cascade for Separating a Mixture of Isotopes with Impurities

A method is proposed for calculating and optimizing a cascade separating a mixture of isotopes with impurities. The approach developed optimizes a cascade with arbitrary mixture composition and a large difference in the masses of the components being separated. The application of this method to the separation of a binary mixture of uranium isotopes with impurities is examined. The properties of cascades optimized with respect to the criterion of minimum total flux are investigated. Their correspondence to R cascades with symmetric steps is analyzed. A comparison is made with optimal cascades with no impurities. It is shown that multicomponent potentials can be used to estimate the separation efficiency for a binary mixture of isotopes with an impurity.__________Translated from Atomnaya Energiya, Vol. 98, No. 4, pp. 293–300, April, 2005.     

Activation Response of Martensitic Steels

A hypothetical martensitic steel has been compositionally designed in order to optimize both metallurgical and reduced activation properties. When compared with two other martensitic steels, its activation characteristics are shown to be superior for all activation indices examined. However, these excellent properties are found to be due to the assumed absence of deleterious tramp impurities. When limiting impurity concentrations are determined for the hypothetical steel, they are found to be extremely stringent, and wholly unachievable using industrial scale production methods. It is concluded that only slight improvements can be made to currently available low activation martensitic steels to reduce residual activity responses further.     

Experience with the fabrication of thick-walled forgings of X 5 CrNi 13 4 for the primary circuit of pressurized water reactors

Parts of the main coolant pump which formerly had been made out of austenitic steels are now produced from the corrosion resistant ferritic steel X 5 CrNi 13 4. This steel has the thermal expansion of ferritic steels, high strength and toughness properties, and allows an ultrasonic examination without problems. The production of forgings with wall thicknesses up to 705 mm in the heat treatment contour is described. Also an evaluation of chemical, mechanical and physical properties obtained from 16 heats of X 5 CrNi 13 4 is given. These results clearly indicate that this steel is recommendable for further application.     

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.     

Mechanical and thermomechanical properties of commercially pure chromium and chromium alloys

The mechanical and thermal properties of commercially pure chromium and the chromium-based alloys Cr–5Fe–1Y₂O₃ and Cr–44Fe–5Al–0.3Ti–0.5Y₂O₃ have been investigated in order to determine the thermal stress factor of these materials and to assess their capability to withstand high-thermal loads in fusion applications. Especially the alloy Cr–5Fe–1Y₂O₃ combines sufficient mechanical strength at temperatures up to 1000 °C, high-thermal conductivity and a low-thermal expansion coefficient to yield the lowest thermal stress factor of all metallic candidate materials for first wall and blanket applications. The high-ductile-to-brittle transition temperature may lead to a rather high value for the lower operation-temperature limit.     

Ferrite formation in neutron-irradiated austenitic stainless steel

Magnetic measurements were carried out on type 316, 321 and three modified heats of 316 austenitic stainless steels that had been irradiated to high fluences (1 ? 8 × 10²²n/cm², E > 0.1 MeV) in EBR-II at temperatures ranging from 450–700°C. Most of the specimens showed increases of magnetization after exposure to the reactor environment that can be attributed to formation of numerous small ferrite particles. The amount of ferrite formed during irradiation is a function of alloy composition as well as irradiation temperature and fluence. Specimens with low molybdenum concentrations had a greater ferrite content than specimens with the normal molybdenum content of type 316 stainless steel. A modified heat of type 316 with 0.23 wt% Ti had lower levels of ferrite under given irradiation conditions than the other heats. Some particles with diffraction patterns corresponding to the ferrite phase were found in an irradiated type 321 stainless specimen, but none were observed in the type 316 stainless specimens.     

ADS散裂中子源材料不锈钢和钨的辐照效应(英文)

研究了80 MeV碳或85 MeV氟离子辐照在国产改进型316L不锈钢、普通不锈钢和钨中产生的辐照效应。实验结果表明,不锈钢的抗辐照性能比钨的好;它们中,国产改进型316L不锈钢具有最好的抗辐照性能。选用不锈钢做ADS散裂中子源的束窗等材料是较好的选择,采用国产改进型316L不锈钢是最佳的选择。     

Impurity Transport in a Simulated Gas Target Divertor

Future generation fusion reactors and tokamaks will require dissipative divertors to handle the high particle and heat loads leaving the core plasma (100–400 MW/m² in ITER). A radiative divertor is proposed as a possible scenario, utilizing a hydrogen target gas to disperse the plasma momentum and trace impurity radiation to dissipate the plasma heat flux. Introducing an impurity into the target hydrogen gas enhances the radiative power loss but may lead to a significant impurity backflow to the main plasma. Thus, impurity flow control represents a crucial design concern. Such impurity flows are studied experimentally in this thesis. The PISCES-A linear plasma device (n 3 × 10¹⁹ m^–3, kT _e 20 eV) has been used to simulate a gas target divertor. To study the transport of impurities, a trace amount of impurity gas (i.e., neon and argon) is puffed near the target plate along with the hydrogen gas. Varying the hydrogen gas puffing rate permits us to study the effects of various background plasma conditions on the transport of impurities. A 1-1/2-D fluid code has been developed to solve the continuity and momentum equations for a neutral and singly ionized impurity in a hydrogen background plasma. The results indicate an axial reduction in the impurity concentration upstream from the impurity puffing source. Impurity entrainment is more effective for higher hydrogen target pressures (and for higher hydrogen plasma densities). However, if there is a reversal of the background plasma flow, impurity particles can propagate past the plasma flow reversal point and are then no longer entrained.     

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.     

Passive safety in a helium-cooled tritium breeding blanket with steel structure

Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accidents such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This paper proposes a new helium-cooled, tritium breeding blanket concept which uses a metallic structure, and which performs significantly better during such accidents than related designs. The proposed blanket uses modified, reduced-activation HT-9 steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m². This concept uses novel features such as: (1) a beryllium-joint design which allows beryllium to be used to conduct heat away from the first wall, while accommodating swelling of the beryllium, and (2) a shield cooled by naturally circulating water. These features help the blanket passively withstand a worst-case undercooling accident scenario.Supported by a USDOE Magnetic Fusion Energy Technology Fellowship.     

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。