A.F. Rowcliffe

Research and development on vanadium alloys for fusion applications

The current status of research and development on unirradiated and irradiated V-Cr-Ti alloys intended for fusion reactor structural applications is reviewed, with particular emphasis on the flow and fracture behavior of neutron-irradiated vanadium alloys. Recent progress on fabrication, joining, oxidation behavior, and the development of insulator coatings is also summarized. Fabrication of large (>500 kg) heats of V-4Cr-4Ti with properties similar to previous small laboratory heats has now been demonstrated. Impressive advances in the joining of thick sections of vanadium alloys using GTA and electron beam welds have been achieved in the past two years, although further improvements are still needed.

Austenitic stainless steels and high strength copper alloys for fusion components

Abstract An austenitic stainless steel (316LN), an oxide-dispersion-strengthened copper alloy (GlidCop Al25), and a precipitation-hardened copper alloy (Cu–Cr–Zr) are the primary structural materials for the ITER first wall/blanket and divertor systems. While there is a long experience of operating 316LN stainless steel in nuclear environments, there is no prior experience with the copper alloys in neutron environments. The ITER first wall (FW) consists of a stainless steel shield with a copper alloy heat sink bonded by hot isostatic pressing (HIP). The introduction of bi-layer structural material represents a new materials engineering challenge; the behavior of the bi-layer is determined by the properties of the individual components and by the nature of the bond interface. The development of the radiation damage microstructure in both classes of materials is summarized and the effects of radiation on deformation and fracture behavior are considered. The initial data on the mechanical testing of bi-layers indicate that the effectiveness of GlidCop Al25 as a FW heat sink material is compromised by its strongly anisotropic fracture toughness and poor resistance to crack growth in a direction parallel to the bi-layer interface.

Effects of Si and Ti on the phase stability and swelling behavior of AISI 316 stainless steel

Abstract The swelling behavior of neutron irradiated stainless steels is strongly influenced by solute segregation and precipitation phenomena. The extent to which in-reactor swelling behavior may be simulated by heavy ion irradiation depends upon the extent to which in-reactor phase changes are reproduced; this question is addressed by comparing the precipitation behavior under neutron irradiation with behavior during 4 MeV Ni ion irradiation for AISI 316 stainless steel and a related stainless steel containing additions of titanium and silicon. The results are discussed qualitatively in terms of the effects of damage rate on solute segregation and the effects of displacement cascades on the dissolution of particles. It is shown that the partitioning of elements into various phases during irradiation is not a sufficient condition for the initiation of swelling in stainless steels modified with silicon and titanium. It is also necessary for helium to be generated simultaneously with the breakdown of the matrix into various phases; it is believed that helium trapping at the growing particle-matrix interface is responsible for the observed physical association between voids and precipitates.

High temperature radiation damage phenomena in complex alloys

Abstract The swelling behavior of structural alloys such as the austenitic stainless steels is closely linked to precipitation behavior during irradiation. Recent experimental evidence is reviewed which illustrates the complexity of this interaction. It is suggested that processes associated with the trapping of helium atoms and point defects at particle-matrix interfaces and the inhibition of climb by precipitation are important factors in determining swelling behavior. These phenomena together with the subject of phase stability in a radiation environment represent important topics for further theoretical and experimental work.

Fracture toughness and tensile behavior of ferritic-martensitic steels irradiated at low temperatures

Abstract Disk compact tension and sheet tensile specimens of the ferritic-martensitic steels F82H and Sandvik HT-9 were irradiated in the High Flux Isotope Reactor (HFIR) at 90°C and 250°C to neutron doses of 1.5–2.5 dpa. For both steels, radiation hardening was accompanied by a reduction in strain hardening capacity (SHC). When combined with other literature data it is apparent that severe loss of SHC occurs in F82H for irradiation temperatures below ∼400°C and in HT-9 for irradiation temperatures below ∼250°C. For both alloys, severe loss of SHC does not correlate with brittle behavior during fracture toughness testing.

Irradiation performance of stainless steels for ITER application

Abstract The proposed normal operating temperature range for the ITER first wall/shield structure (100–250°C) is below the temperature regimes for void swelling (400–600°C) and for grain boundary embrittlement (500–700°C). However, the neutron doses for the basic performance phase (3–4 dpa) and the extended performance phase (20–30 dpa) are such that large changes in yield strength, deformation mode, and strain hardening capacity will be encountered which could significantly affect fracture properties. Yield strength increases rapidly with dose in the 60–300°C regime with the increase tending to saturate at 1–3 dpa. Under certain conditions, radiation hardening is accompanied by changes in the stress-strain relationship with the appearance of an initial yield drop and a significant reduction in strain hardening capacity. This paper reviews the low temperature (

Progress in vanadium alloy development for fusion applications

Abstract Vanadium alloys have been identified as a leading candidate low-activation structural material for fusion first-wall blanket applications. Candidate vanadium alloys exhibit favorable safety and environmental characteristics, good fabricability, high temperature and heat load capability, good compatibility with liquid metals and resistance to irradiation damage. The focus of the vanadium alloy development program has been on the vanadium-chromium-titanium (0–15%Cr, 1–20%Ti) alloy system. Investigations include effects of minor alloy elements such as Si, Al and Y and substitution of iron for chromium in the ternary alloy. A V-4Cr-4Ti alloy is currently regarded as the reference alloy. Significant progress has been made in the development of vanadium alloys for fusion applications. Two production-scale heats (500 and 1200 kg) of the V-4Cr-4Ti alloys have been produced with controlled levels of impurities. The baseline properties of the 500 kg heat are similar to those of the previous laboratory-scale heats. Additional data have been obtained on baseline tensile and fracture properties. Results obtained on several heats with minor variations in composition indicate high uniform and total elongation of these alloys at temperature from RT to 700°C. Results obtained to date indicate that the V-Cr-Ti alloys are resistant to swelling and embrittlement after exposure to relatively high neutron fluences at temperatures of 400–600°C. The properties are not significantly different when modest amounts of helium are generated during neutron irradiation by the dynamic helium charging experiment method. However, recent results have indicated that these alloys are susceptible to irradiation embrittlement at lower temperatures. Additional irradiation experiments are in progress to investigate these effects at temperatures of 200–400°C. This paper presents an update on the experimental results on candidate low activation vanadium alloys.

Swelling behavior of austenitic stainless steels in a spectrally tailored reactor experiment: Implications for near-term fusion machines

Abstract Current designs for engineering test reactors such as the International Thermonuclear Experimental Reactor propose to use an austenitic stainless steel for the first wall. Most of the available swelling data have been derived from neutron-irradiation experiments in which helium generation rates are very low (fast breeder reactors) or very high (mixed spectrum reactors). Recently a spectrally tailored experiment was concluded in the Oak Ridge Research Reactor in which the helium generation rate and damage rate were maintained at values typical of a fusion reactor operating at ~ 1 MW/m 2 . It was found that the swelling behavior of a titanium-modified stainless steel (PCA) in both the cold-worked and solution-annealed conditions differed significantly from the behavior observed in earlier experiments in which the He/dpa ratio was either ~ 0.5 or ~ 50. The results suggest that there is a strong dependence of microstructural evolution on the He/dpa ratio. The data are shown to be consistent with earlier theoretical predictions of swelling behavior that is a non-monotonic function of the He/dpa ratio. Finally both the present data set and a larger collection of low-temperature swelling data are discussed in the context of near-term machines.

Austenitic stainless steels with improved resistance to radiation-induced swelling☆

Stainless steel specimens were bombarded with Ni ions at elevated temperatures to simulate fast neutron damage. Results show that type 316 stainless steel with additions of silicon and titanium exhibits low swelling over the entire swelling temperature range under high-dose nickel-ion bombardment. Neutron irradiation data on commercial alloys and ion data on high-purity alloys also indicate that the most effective suppression of swelling is achieved with combinations of silicon and titanium. It is suggested that suppression of swelling by alloying with silicon and titanium may be effective over a range of nickel and chromium base composition levels and will provide the basis for the development of low-swelling alloys that are technologically similar to type 316 stainless steel. It should be noted that the influence of silicon and titanium on swelling is likely to depend strongly on the concentrations of other elements such as carbon, oxygen, and nitrogen and on the extent to which silicon and titanium partition to various phases.

Structural materials for ITER in-vessel component design

Copyright (c) 1996 Elsevier Science B.V. All rights reserved. The materials proposed for ITER in-vessel components have to exhibit adequate performance for the operating lifetime of the reactor or for specified replacement intervals. Estimates show that maximum irradiation dose to be up to 5−7 dpa (for 1 MWa/m 2 in the basic performance phase (BPPrr within a temperature range from 20 to 300°C. Austenitic SS 316LN-ITER Grade was defined as a reference option for the vacuum vessel, blanket, primary wall, pipe lines and divertor body. Conventional technologies and mill products are proposed for blanket, back plate and manifold manufacturing. HIPing is proposed as a reference manufacturing method for the primary wall and blanket and as an option for the divertor body. The existing data show that mechanical properties of HIPed SS are no worse than those of forged 316LN SS. Irradiation will result in property changes. Minimum ductility has been observed after irradiation in an approximate temperature range between 250 and 350°C, for doses of 5−10 dpa. In spite of radiation-induced changes in tensile deformation behavior, the fracture remains ductile. Irradiation assisted corrosion cracking is a concern for high doses of irradiation and at high temperatures. Re-welding is one of the critical issues because of the need to replace failed components. It is also being considered for the replacement of shielding blanket modules by breeding modules after the BPP. Estimates of radiation damage at the locations for re-welding show that the dose will not exceed 0.05 dpa (with He generation of 1 appmr for the manifold and 0.01 dpa (with He generation 0.1 appmr for the back plate for the BPP of ITER operation. Existing experimental data show that these levels will not result in property changes for SS; however, neutron irradiation and He generation promote crack formation in the heat affected zone during welding. Cu based alloys, DS-Cu (Glidcop Al25r and PH-Cu (Cu−Cr−Zr bronzer are proposed as a structural materials for high heat flux components of limiter, baffle, divertor and primary wall. Irradiation significantly changes the mechanical properties, and the electrical and thermal conductivity of these alloys. The ductility of high strength Cu alloys is reduced at relatively low doses (<0.2 dpar for irradiation temperature s<150°C. For higher doses of irradiation it remains at the low (saturatedr level. This effect is exhibited by both DS-Cu and PH-Cu alloys. For higher temperatures of irradiation, an increase of ductility and decrease of strength are observed resulting from radiation-induced microstructural instabilities. The softening temperature for Cu−Cr−Zr alloys is in the range 230−250°C; the corresponding temperature for DS Cu alloys is ca. 300−400°C.