Helium-implanted CLAM steel and evolutionary behavior of defects investigated by positron-annihilation spectroscopy

China Low Activation Martensitic (CLAM) steel has been chosen as the primary candidate structural material for the first wall/blanket for fusion reactor. The excessive helium irradiation induced damage of CLAM steel at high temperatures and the evolution of defects were investigated in this paper. The samples were homogeneously implanted with 1e + 17 ions/cm² and 100 keV of helium at room temperature, 473, 673, and 873 K. Irradiation induced damage of CLAM steel and the annealing behavior of defects were probed by slow positron beam Doppler broadening technique. Helium implantation produced a large number of vacancy-type defects which bound with helium and formed helium–vacancy complexes. Target atoms’ displacement capacity was strengthened with rising irradiation temperatures, so the S parameter increased with increasing irradiation temperatures, and helium–vacancy complexes were main defects after helium implantation at damage layers. Helium bubbles would be unstable and the desorption of helium bubbles would promote the density of defects above 673 K. By analyzing the curves of S–W and annealing tests of irradiated specimen, it suggested that there werenot only one type of defect in damage layers. Though helium–vacancy complexes were primary defects after helium implanted, introducing excessive helium might also generated other point defects or dislocation loops in the material.     

Ion-irradiation-induced damage of steels characterized by means of nanoindentation

Self-ion irradiation was used to simulate the damage caused by fast neutrons in the austenitic stainless steel SS 304 SA, the ferritic/martensitic steel Eurofer’97 and a Fe–9 at.%Cr model alloy. The irradiation-induced hardness change in the damage layer was evaluated by means of nanoindentation. Three-step irradiations were performed at room temperature and 300 °C up to 1 and 10 dpa. An irradiation-induced hardness change was shown for all materials. No influence of irradiation temperature could be resolved. Irradiation-induced hardening exhibits different fluence dependencies in Eurofer’97 and Fe–9 at.%Cr. While the data indicate a saturation-like behaviour for Fe–9 at.%Cr, an increase of hardness with fluence up to 10 dpa was found for Eurofer’97.     

Measurement and analysis of SHCCT diagram for CLAM steel

China Low Activation Martensitic (CLAM) steel is a leading candidate material for construction of the Chinese fusion reactor Test Blanket Module. The Simulated HAZ Continuous Cooling Transformation (SHCCT) diagram is developed via physical simulation, and the effects of thermal history on the microstructure and mechanical properties of the weld coarse-grain heat-affected zone (CGHAZ) in CLAM steel are evaluated. The results of thermal cycle simulation show that grain size increases and hardness decreases gradually with increasing heat input. Under certain conditions, especially when cooling times from 800 °C to 500 °C (T8/5) are larger than 136 s, delta ferrite may form which is deleterious for the TBM application. The amounts of delta ferrite are given under different T8/5. A SHCCT diagram of CLAM steel is developed using dilatometry and it predicts the AC1, AC3 and the Ms temperatures. With decreased cooling rate (larger T8/5), martensite laths widen and carbide precipitates grow. The results indicate that welding heat input should be taken into consideration and controlled in practical CLAM steel welding process applications.     

Evaluation of soft error rates using nuclear probes in bulk and SOI SRAMs with a technology node of 90 nm

The difference of soft error rates (SERs) in conventional bulk Si and silicon-on-insulator (SOI) static random access memories (SRAMs) with a technology node of 90 nm has been investigated by helium ion probes with energies ranging from 0.8 to 6.0 MeV and a dose of 75 ions/μm². The SERs in the SOI SRAM were also investigated by oxygen ion probes with energies ranging from 9.0 to 18.0 MeV and doses of 0.14–0.76 ions/μm². The soft error in the bulk and SOI SRAMs occurred by helium ion irradiation with energies at and above 1.95 and 2.10 MeV, respectively. The SER in the bulk SRAM saturated with ion energies at and above 2.5 MeV. The SER in the SOI SRAM became the highest by helium ion irradiation at 2.5 MeV and drastically decreased with increasing the ion energies above 2.5 MeV, in which helium ions at this energy range generated the maximum amount of excess charge carriers in a SOI body. The soft errors occurred by helium ions were induced by a floating body effect due to generated excess charge carriers in the channel regions. The soft error occurred by oxygen ion irradiation with energies at and above 10.5 MeV in the SOI SRAM. The SER in the SOI SRAM gradually increased with energies from 10.5 to 13.5 MeV and saturated at 18 MeV, in which the amount of charge carriers induced by oxygen ions in this energy range gradually increased. The computer calculation indicated that the oxygen ions with energies above 13.0 MeV generated more excess charge carriers than the critical charge of the 90 nm node SOI SRAM with the designed over-layer thickness. The soft errors, occurred by oxygen ions with energies at and below 12.5 MeV, were induced by a floating body effect due to the generated excess charge carriers in the channel regions and those with energies at and above 13.0 MeV were induced by both the floating body effect and generated excess carriers. The difference of the threshold energy of the oxygen ions between the experiment and the computer calculation might be due to the difference between the designed and real structures.     

Synergistic effect of helium and hydrogen for bubble swelling in reduced-activation ferritic/martensitic steel under sequential helium and hydrogen irradiation at different temperatures

In order to investigate the synergistic effect of helium and hydrogen on swelling in reduced-activation ferritic/martensitic (RAFM) steel, specimens were separately irradiated by single He⁺ beam and sequential He⁺ and H⁺ beams at different temperatures from 250 to 650 °C. Transmission electron microscope observation showed that implantation of hydrogen into the specimens pre-irradiated by helium can result in obvious enhancement of bubble size and swelling rate which can be regarded as a consequence of hydrogen being trapped by helium bubbles. But when temperature increased, Ostwald ripening mechanism would become dominant, besides, too large a bubble could become mobile and swallow many tiny bubbles on their way moving, reducing bubble number density. And these effects were most remarkable at 450 °C which was the peak bubble swelling temperature for RAMF steel. When temperature was high enough, say above 450, point defects would become mobile and annihilate at dislocations or surface. As a consequence, helium could no longer effectively diffuse and clustering in materials and bubble formation was suppressed. When temperature was above 500, helium bubbles would become unstable and decompose or migrate out of surface. Finally no bubble was observed at 650 °C.     

He implantation induced microstructure- and hardness-modification of the intermetallic γ-TiAl

TiAl is a well known high temperature material with good creep properties. It is investigated as a potential structural material for Generation IV high temperature gas cooled nuclear reactors. The tests are performed with the ABB-2 (Ti-rich TiAl with 2 at.% W) developed by ASEA Brown Boveri Ltd. (ABB). Thin samples are irradiated throughout with 24 MeV ⁴He²⁺ ions; the irradiated material is then investigated towards its microstructure and its hardness. The microstructure is studied by transmission electron microscopy and the hardness is investigated using a micro-hardness tester and a nano-indenter. Different effects can be identified. From room to moderate irradiation temperatures, the radiation induced hardening of the material slowly vanishes until the material completely recovers at about 943 K. Beyond this temperature, He-bubble formation seems to harden the material again, until beyond 1200 K a steep increase in hardening is detected. This effect can be correlated with bubbles being identified in the micrographs. The results are consistent and give strong indications to a microstructural development as a function of temperature.     

Irradiation dose and temperature dependence of fracture toughness in high dose HT9 steel from the fuel duct of FFTF

To expand the knowledge base for fast reactor core materials, fracture toughness has been evaluated for high dose HT9 steel using miniature disk compact tension (DCT) specimens. The HT9 steel DCT specimens were machined from the ACO-3 fuel duct of the Fast Flux Test Facility (FFTF), which achieved high doses in the range of 3–148 dpa at 378–504 °C. The static fracture resistance (J-R) tests have been performed in a servohydraulic testing machine in vacuum at selected temperatures including room temperature, 200 °C, and each irradiation temperature. Brittle fracture with a low toughness less than 50 MPa √m occurred in room temperature tests when irradiation temperature was below 400 °C, while ductile fracture with stable crack growth was observed when irradiation temperature was higher. No fracture toughness less than 100 MPa √m was measured when the irradiation temperature was above 430 °C. It was shown that the influence of irradiation temperature was dominant in fracture toughness while the irradiation dose has only limited influence over the wide dose range 3–148 dpa. A slow decrease of fracture toughness with test temperature above room temperature was observed for the nonirradiated and high temperature (>430 °C) irradiation cases, which indicates that the ductile–brittle transition temperatures (DBTTs) in those conditions are lower than room temperature. A comparison with the collection of existing data confirmed the dominance of irradiation temperature in the fracture toughness of HT9 steels.     

Hydrogen implantation defects in MgO

Deuterium and hydrogen ions with an energy of 15 keV have been implanted in virgin MgO (1 0 0) single crystals and in single crystals containing helium implantation generated microcavities. Doses were varied from 2 × 10¹⁵ to 2 × 10¹⁶ cm⁻². The samples were annealed from room temperature to 950 K. The defects produced by hydrogen and the trapping of hydrogen at the defects were monitored by photon absorption and positron beam analysis. With this novel technique a depth distribution of defects can be determined for implantation depths from 0 to 2000 nm. The technique is very sensitive for vacancy and vacancy clusters, i.e. sites with low electron density. After 950 K annealing microcavities were observed for the 2 × 10¹⁶ cm⁻² dose but not for the 10 times lower dose. During annealing up to 750 K point defects are mobile but the defect clusters remain small and filled with hydrogen. In samples which contain already microcavities, point defects and deuterium from the deuterium irradiation are accumulated by the microcavities.     

Deuterium and helium retentions of V–4Cr–4Ti alloy used as first wall of breeding blanket in a fusion reactor

The deuterium and helium retention properties of V–4Cr–4Ti alloy were investigated by thermal desorption spectroscopy (TDS). Ion energies of deuterium and helium were taken at 1.7 and 5 keV, respectively. The retained amount of deuterium in the sample irradiated at 380 K increased with the ion fluence and was not saturated to fluence of up to 1 × 10²³ D/m². For the irradiation at 773 K, 0.1% of implanted deuterium was retained at the highest fluence. For the helium ion irradiation at room temperature, three groups of desorption peaks appeared at around 500, 850, and 1200 K in the TDS spectrum. In the lower fluence region (<1 × 10²¹ He/m²), the retained helium desorbed mainly at around 1200 K. With increasing fluence, the amount desorbed at 500 K increased. Total amount of retained helium in the samples saturated at fluence up to 5 × 10²¹ He/m² and saturation level was 2.7 × 10²¹ He/m².     

Compatibility of CLAM steel weldments with static LiPb alloy at 550 °C

CLAM steel is considered as a structural material to be used in the Test Blanket Module as a barrier or blanket adjacent to liquid LiPb in fusion reactors. In this paper, CLAM steel is welded by tungsten inert gas (TIG) welding, and the compatibility of the weldment with liquid LiPb is tested. Specimens were corroded in static liquid LiPb, with corrosion times of 500 h and 1000 h, at 550 °C, and the corresponding weight losses are 0.272 mg/cm² and 0.403 mg/cm² respectively. Also the corrosion rate decreases with increased corrosion time. In the as-welded condition, corrosion resistance of the weld zone is higher than that of the HAZ (Heat Affected Zone). Likely, thick martensite lath and large residual stresses at the welding zone result in higher corrosion rates. The compatibility of CLAM steel weld joints with high temperature liquid LiPb can be improved to some extent through a post-weld tempering process. The surface of the as-welded CLAM steel is uniformly corroded and the concentration of Cr on the surface decreases by about 50% after corrosion. Penetration of LiPb into the matrix is observed for neither the as-welded nor the as-tempered conditions. Influenced by thick martensite lath and large residual stresses, the welded area, especially the weld zone, is easily corroded, therefore it is of primary importance to protect the welded area in the solid blanket of the fusion reactor.     

Behaviour of implanted oxygen and nitrogen in halogen lamp annealed silicon

The behaviour of oxygen and nitrogen sequentially implanted into silicon at an energy of 100 and 175 keV with total doses of 5 × 10¹⁶ and 2 × 10¹⁷ cm⁻² is reported. The implantation sequence and energies were varied. In all cases the wafer temperature during implantation was maintained at 300°C. The samples were pulse annealed in using a halogen lamp to achieve temperatures of 1100°C and 1200°C for 2–60 s. The samples were analysed using FTIR, XRD and SIMS. For a total dose of 5 × 10¹⁶ cm⁻² no mutual redistribution of oxygen and nitrogen occurs during annealing. The impurity atoms (O and N) are found to bind to radiation defects. Annealing does not lead to any detectable new phase formation but does cause dissociation of oxygen–vacancy complexes and loss of oxygen from the bulk. For a dose of 2 × 10¹⁷ cm⁻² phase formation occurs during implantation. The mutual redistribution of the impurity atoms during annealing occurs but only if oxygen is implanted deeper than nitrogen. This redistribution is greatest if nitrogen is implanted prior to oxygen.     

The effect of displacement damage on deuterium retention in ITER-grade tungsten exposed to low-energy,high-flux pure and helium-seeded deuterium plasmas

Samples prepared from polycrystalline ITER-grade tungsten were damaged by irradiation with 20 MeV W ions at room temperature to a fluence of 1.4 × 10¹⁸ W/m². Due to the irradiation, displacement damage peaked near the end-of-range, 1.35 μm beneath the surface, at 0.89 displacements per atom. The damaged as well as undamaged W samples were then exposed to low-energy, high-flux (10²² D/m² s) pure D and helium-seeded D plasmas to an ion fluence of 3 × 10²⁶ D/m² at various temperatures. Trapping of deuterium was examined by the D(³He,p)⁴He nuclear reaction at ³He energies varied from 0.69 to 4.0 MeV allowing determination of the D concentration at depths up to 6 μm. It has been found that (i) addition of 10% helium ions into the D plasma at exposure temperatures of 440–650 K significantly reduces the D concentration at depths of 0.5–6 μm compared to that for the pure plasma exposure; (ii) generation of the W-ion-induced displacement damage significantly increases the D concentration at depths up to 2 μm (i.e., in the damage zone) under subsequent exposures to both pure D and D–He plasmas.     

Weldability of neutron irradiated austenitic stainless steels

Degradation of weldability in neutron irradiated austenitic stainless steel is an important issue to be addressed in the planning of proactive maintenance of light water reactor core internals. In this work, samples selected from reactor internal components which had been irradiated to fluence from 8.5 × 10²² to 1.4 × 10²⁶ n/m² (E > 1 MeV) corresponding to helium content from 0.11 to 103 appm, respectively, were subjected to tungsten inert gas arc (TIG) welding with heat input ranged 0.6–16 kJ/cm. The weld defects were characterized by penetrant test and cross-sectional metallography. The integrity of the weld was better when there were less helium and at lower heat input. Tensile properties of weld joint containing 0.6 appm of helium fulfilled the requirement for unirradiated base metal. Repeated thermal cycles were found to be very hazardous. The results showed the combination of material helium content and weld heat input where materials can be welded with little concern to invite cracking. Also, the importance of using properly selected welding procedures to minimize thermal cycling was recognized.     

Tensile stress–strain and work hardening behaviour of P9 steel for wrapper application in sodium cooled fast reactors

Tensile flow behaviour of P9 steel with different silicon content has been examined in the framework of Hollomon, Ludwik, Swift, Ludwigson and Voce relationships for a wide temperature range (300–873 K) at a strain rate of 1.3 × 10^?3 s^?1. Ludwigson equation described true stress (σ)–true plastic strain (ε) data most accurately in the range 300–723 K. At high temperatures (773–873 K), Ludwigson equation reduces to Hollomon equation. The variations of instantaneous work hardening rate (θ = dσ/dε) and θσ with stress indicated two-stage work hardening behaviour. True stress–true plastic strain, flow parameters, θ vs. σ and θσ vs. σ with respect to temperature exhibited three distinct temperature regimes and displayed anomalous behaviour due to dynamic strain ageing at intermediate temperatures. Rapid decrease in flow stress and flow parameters, and rapid shift in θ–σ and θσ–σ towards lower stresses with increase in temperature indicated dominance of dynamic recovery at high temperatures.     

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.     

EPR study of defects produced by MeV Ag ion implantation into silicon

The damage produced by implanting (1 1 1) Si wafers with 4 MeV Ag ions at implantation temperatures of 210, 350 and 400 K has been investigated by electron paramagnetic resonance as a function of implantation fluence in the range 5 × 10¹²–2 × 10¹⁵ Ag cm⁻². For each implantation temperature, at low ion fluences the EPR spectra show the presence of the point defect centres Si-P3 (neutral 4-vacancy) and Si-P6 (di-interstitial) as well the so-called Σ defect complexes. As the implantation fluence is raised the population of P3 centres goes through a maximum while the Σ centre resonance is gradually replaced by the spectrum of the well-known Si-D centre of a-Si. For implantation at 210 K the total Σ+D centre concentration increases linearly with implantation fluence up to the point at which an amorphous layer is formed; however raising the implantation temperature causes the dependence of the Σ+D concentration on implantation fluence to become increasingly sublinear with the result that the production of a given level of damage requires a larger implantation fluence. The results are discussed in the context of a previous study of the implantation damage in the same samples by optical reflectivity depth profiling [Mat. Res. Soc. Symp. Proc. 540 (1999) 31].     

Effect of keV ion irradiation on mechanical properties of hydrogenated amorphous silicon

The susceptibility of mechanical properties of hydrogenated amorphous silicon (a-Si:H) to the implantation-enhanced disorder has been studied with the aim to extend the application field of this material in the technology of micro-electromechanical systems. Effect of keV ion irradiation on the elastic modulus, E, of hardness, H, and of root-mean-squared roughness to silicon ion implantation has been determined. The mechanical properties were evaluated by nanoindentation testing. E of 119 GPa and H of 12.3 GPa were determined for the as-prepared a-Si:H film. The implantation of silicon ions leads to a decrease in E and H, evaluated for a series of the implantation fluences in the range of 1.0 × 10¹³–5.0 × 10¹⁶ cm^?2. Surface smoothing has been observed at high fluences and low ion energy of 18 keV, suggesting that ion beam may be used as a tool to reduce the roughness of the a-Si:H surface, while keeping intact the mechanical properties inside the film. The conducted experiments show that it is possible to prepare a-Si:H films with hardness and smoothness comparable to crystalline silicon.     

Phase transformation induced by ion implantation in cubic stabilized zirconia

Cubic yttria-stabilized zirconia possesses a high stability against radiation. No amorphization of this material has been observed, even at high ion fluences leading to the production of a large amount of defects. Nevertheless irradiation with energetic particles may induce microstructural evolutions and phase transformations. In the present paper we demonstrate that a cubic-to-rhombohedral phase transformation occurs in yttria-stabilized zirconia implanted with He ions. This transformation consists in a rhombohedral deformation of the cubic cell along the 〈1 1 1〉 directions due to residual stresses induced by implantation.     

Effect of He-pre-injection on dislocation loop formation and irradiation-induced segregation of Fe–12%Cr–15%Mn austenitic steel

The effect of He-injection on irradiation-induced segregation of aging treated Fe–12%Cr–15%Mn austenitic steels, which are candidate materials as the reduced radio-activation of structure material for nuclear and/or fusion reactors was investigated by using the 1250 kV high voltage electron microscope (HVEM) connected with an ion accelerator. The Fe–Mn–Cr steel has been irradiated at 573 K by three irradiation modes of single electron-beam irradiation, electron-beam irradiation after He-injection and electron/He⁺-ion dual-beam irradiation in a HVEM. Irradiation-induced segregation analyses were carried out by an energy dispersive X-ray analyzer (EDX) in a 200 kV FE-TEM with beam diameter of about 0.5 nm. Dislocation loops with strain contrast were formed during irradiation and the loop numbers density increased rapidly with irradiation dose for He-pre-injected specimens. Voids were not observed after irradiations with three irradiation modes up to 5.4 dpa at 573 K. Irradiation-induced segregations of Cr and Mn near grain boundary were observed in each irradiation condition, but the amounts of Mn segregation decreased in the cases of electron/He⁺-ion dual-beam irradiation compared with single electron-beam and electron-beam irradiation after He-injection conditions.     

The use of NRA to study thermal diffusion of helium in (U, Pu)O2

A lot of work has been already done on helium atomic diffusion in UO₂ samples, but information is still lacking about the fate of helium in high level damaged UOX and MOX matrices and more precisely their intrinsic evolutions under alpha self irradiation in disposal/storage conditions.The present study deals with helium atomic diffusion in actinide doped samples versus damage level. The presently used samples allow a disposal simulation of about 100 years of a UOX spent fuel with a 60 MW d kg^?1 burnup or a storage simulation of a MOX spent fuel with a 47.5 MW d kg^?1 burnup.For the first time, nuclear reaction analysis of radioactive samples has been performed in order to obtain diffusion coefficients of helium in (U, Pu)O₂. Samples were implanted with ³He⁺ and then annealed at temperatures ranging from 1123 K to 1273 K. The evolution of the ³He depth profiles was studied by the mean of the non-resonant reaction: ³He(d, p)⁴He. Using the SIMNRA software and the second Fick’s law, thermal diffusion coefficients have been measured and compared to the ³He thermal diffusion coefficients in UO₂ found in the literature.