Vacancy-type defects and hardness of helium implanted CLAM steel studied by positron-annihilation spectroscopy and nano-indentation technique

China Low Activation Martensitic (CLAM) steel was implanted with helium up to 1e + 16/cm² at 300–873 K using 140 keV helium ions. Vacancy-type defects induced by implantation were investigated with positron beam Doppler broadening technique, and then nano-hardness measurements were performed to investigate helium-induced hardening effect. He implantation produced a large number of vacancy-type defects in CLAM steel, and the concentration of vacancy-type defects decreased with increasing temperature. Vacancy–helium complexes were main defects at different temperatures. Irradiation induced hardening was observed at all irradiation temperatures, and the peak value of hardness was at 473 K. The result suggested that both vacancy–helium complexes and helium bubbles had contribution to irradiation induced hardening. The decomposition and annihilation of irradiation-induced defects became more and more significant with increasing temperature, which induced the increment of hardness became more and more small.     

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.     

Enhanced radiation tolerance of ultrafine grained Fe–Cr–Ni alloy

The evolutions of microstructure and mechanical properties of Fe–14Cr–16Ni (wt.%) alloy subjected to Helium ion irradiations were investigated. Equal channel angular pressing (ECAP) process was used to significantly reduce the average grain size from 700 μm to 400 nm. At a peak fluence level of 5.5 displacement per atom (dpa), helium bubbles, 0.5–2 nm in diameter, were observed in both coarse-grained (CG) and ultrafine grained (UFG) alloy. The density of He bubbles, dislocation loops, as well as radiation hardening were reduced in the UFG Fe–Cr–Ni alloy comparing to those in its CG counterpart. The results imply that radiation tolerance in bulk metals can be effectively enhanced by refinement of microstructures.     

Thermal conductivity of highly neutron-irradiated beryllium in nuclear fusion reactors

Beryllium will be used as a neutron multiplier in Helium Cooled Pebble Bed (HCPB) DEMO blankets. The beryllium thermal conductivity is determining the maximum pebble bed temperature and, therefore, is very important for blanket design. Different grades of beryllium discs were neutron-irradiated at temperatures between 343 and 673 K and at fluences up to 1.6 × 10²³ cm⁻². At lower irradiation temperatures a significant drop of the beryllium thermal conductivity occurs even after small neutron fluences. With increasing neutron fluence, further moderate decreases of the conductivity are observed. With increasing irradiation temperature, the thermal conductivity further decreases. If the thermal conductivity of the irradiated beryllium is known, the conductivity of irradiated beryllium pebble beds can be assessed using the model suggested in this study.     

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.     

Investigation of defects in actinide-doped UO2 by positron annihilation spectroscopy

The nuclearization and validation of a new positron annihilation lifetime spectroscopy (PALS) system was ideally used to investigate vacancy defects generated by α self-irradiation in the UO₂ matrix of several plutonium-doped samples. The damage levels studied ranged from 0 to 0.3 dpa. This study validated the operational protocols for actinide-doped materials. A lattice lifetime of about 170–180 ps was determined for the undoped UO₂ matrix, which is consistent with the values reported in the literature. Alpha self-irradiation damage systematically increases the mean positron lifetime, resulting in a difference of 133 ps for a damage level of 0.3 dpa. Even at low damage values, a positron trapping site appears that corresponds to point defects involving an uranium vacancy, with a specific lifetime of about 310 ps. When annealed at 1373 K, some of these defects coalesce to form larger extended defects. The initial results for actinide-doped UO₂ also confirm the high sensitivity of PALS to the presence of vacancy defects even at low integrated α dose.     

Theoretical study of helium insertion and diffusion in 3C-SiC

Insertion and diffusion of helium in cubic silicon carbide have been investigated by means of density functional theory. The method was assessed by calculating relevant properties for the perfect crystal along with point defect formation energies. Results are consistent with available theoretical and experimental data. Helium insertion energies were calculated to be lower for divacancy and silicon vacancy defects compared to the other mono-vacancies and interstitial sites considered. Migration barriers for helium were determined by using the nudged elastic band method. Calculated activation energies for migration in and around vacancies (silicon vacancy, carbon vacancy or divacancy) range from 0.6 to 1.0 eV. Activation energy for interstitial migration is calculated to be 2.5 eV. Those values are discussed and related to recent experimental activation energies for migration that range from 1.1 [P. Jung, J. Nucl. Mater. 191–194 (1992) 377] to 3.2 eV [E. Oliviero, A. van Veen, A.V. Fedorov, M.F. Beaufort, J.F. Bardot, Nucl. Instrum. Methods Phys. Res. B 186 (2002) 223; E. Oliviero, M.F. Beaufort, J.F. Bardot, A. van Veen, A.V. Fedorov, J. Appl. Phys. 93 (2003) 231], depending on the SiC samples used and on helium implantation conditions.     

Thermal desorption of helium from defects in nickel

Helium atoms, introduced into materials by helium plasma or generated by the (n, α) nuclear reaction, have a strong tendency to accumulate at trapping sites such as vacancy clusters and dislocations. In this paper, the effects of dislocations, single vacancies and vacancy clusters on the retention and desorption of helium atoms in nickel were studied. Low energy (0.1-0.15 keV) helium atoms were implanted in nickel with vacancies or dislocations without causing any displacement damage. He atoms, interstitial-type dislocation loops, and vacancy clusters were also introduced with irradiation damage by 5.0 keV helium ions. Helium thermal desorption peaks from dislocations, helium-vacancy clusters and helium bubbles were obtained by thermal desorption spectroscopy at 940 K, in the range from 900 to 1370 K, and at 1500 K, respectively. In addition, a thermally quasi-stable state was found for helium-vacancy clusters.     

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.     

X-ray diffraction analysis of titanium tritide film during 1600 days

The generation and accumulation of ³He by tritium decay modified the physical and chemical properties of tritides. Here the evolution of lattice defects in long-aged titanium tritide films is investigated by X-ray diffraction and changes in the positions, intensities and line shapes of diffraction peaks have been determined over a period of about 1600 days (>4 years). Texture effects are also observed by biased intensities in standard θ–2θ scans. The results show that the TiT_1.5 film keeps an fcc structure during 1600 days and reveals an hkl-dependent unit-cell expansion and line width broadening which are interpreted in terms of isolated tetrahedral interstitial ³He atoms and isolated bubble growth models by dislocation loop-punching or dislocation dipole expansion combined with Krivoglaz theory. In the first 12 days of aging, isolated tetrahedral interstitial ³He atoms or ³He clusters are formed, then interstitial ³He atoms diffuse into (1 1 1) planes and precipitate into clusters. The spontaneous formation of Frenkel pairs, the self-interstitial atoms produced are built into dislocations resulting in formation platelet bubbles and dislocation dipoles between 12 and 27 days. Above 27 days, multiple stages of ³He bubbles growth appear: (1) between 27 and 85 days platelet helium bubbles growth by dislocation dipoles expansion, (2) between 85 and 231 days the transition from platelet bubbles to sphere bubbles by loop emission, (3) after 231 days sphere bubbles growth by dislocation loop-punching and probably formation of sub-grain boundaries by dislocation rearrangement.     

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.     

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.     

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.     

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.     

DIADDHEM set-up: New IBA facility for studying the helium behavior in nuclear glasses

The immobilization of fission products and minor actinides by vitrification is the reference process for industrial management of high-level radioactive wastes generated from spent fuel reprocessing. The glassy matrix is subjected to radiation damage and radiogenic helium generation due to the alpha decays of minor actinides.A specific experimental study has been conducted to better understand the behavior of helium and its diffusion mechanisms in the borosilicate glass. Helium production is simulated by external irradiation with ³He⁺ ions at a concentration (2 × 10¹⁵ He cm^?2) equivalent to the one obtained after 1000 years of glass storage. He diffusion coefficients as function of temperature are extracted from the evolution of the depth profiles after annealing. The ³He(d, α) ¹H Nuclear Reaction Analysis (NRA) technique is successfully used for in situ low-temperature measurements of depth profiles. Its high depth resolution allows detecting helium mobility at a temperature as low as 250 K and the presence of a trapped helium fraction. The good agreement of our first values of diffusion coefficients with the literature data highlights the relevance of the implantation technique in the study of helium diffusion mechanisms in borosilicate glasses.     

Quantifying helium distribution in dual-ion beam irradiated SiCf/SiC composites by electron energy loss spectroscopy

In this study, we report a method to quantify the helium distribution in the SiC_f/SiC composites, which are used as the first-wall materials of fusion reactor. The helium-bubble formation in Hi-Nicalon Type-S (HNS) was observed in the irradiated SiC_f/SiC composites at a level of 100 dpa and at 800 °C and 1000 °C, respectively. We applied transmission electron microscopy and electron energy loss spectroscopy to investigate the helium-gas-bubbles-formation mechanisms. To simulate the practical first-wall environment of Deuterium–Tritium (D–T) fusion reactor, a dual-ion beam (6 MeV Si³⁺ and 1.13 MeV He⁺) was performed to irradiate the SiC_f/SiC composites. The relationship between the energy shift of He K-edge and the radius of the bubble of the SiC composites was estimated by electron energy loss spectroscopy analysis. The results show that all of the helium atoms irradiated at 1000 °C and formed the bubbles. On the other hand, at 800 °C, only 25.5% of the helium atoms form the helium bubbles. A clear thermal-dependent formation mechanism is found.     

The effect of neutron irradiation dose on vacancy defect accumulation and annealing in pure nickel

In order to investigate the dose dependence of vacancy defect evolution in nickel, specimens of high-purity Ni were neutron-irradiated at ～330 K in the IVV-2M reactor (Russia) to fluencies in the range of 1 × 10²¹–1 × 10²³ n/m² (E > 0.1 MeV) corresponding to displacement dose levels in the range of about 0.0001–0.01 dpa and subsequently stepwise annealed to about 900 K. Ni was characterized both in as-irradiated state as well as after post-irradiation annealing by positron annihilation spectroscopy. The formation of three-dimensional vacancy clusters (3D-VCs) in cascades was observed under neutron irradiation, the concentration of 3D-VCs increases with increasing dose level. 3D-VCs collapse into secondary-type clusters (stacking fault tetrahedra (SFTs), and vacancy loops) during stepwise annealing at 350–450 K. It is shown that the thermal stability of SFTs grow with increasing dose level, probably, it is due to growth of the average SFT size during annealing. The results of annealing experiments on electron-irradiated Ni at 300 K are indicated in the paper, for comparison. We also have briefly discussed the positron response to the SFT-like structures.     

TEM analysis of high temperature annealed W nanostructure surfaces

Helium plasma irradiation and electron heating experiments were conducted using tungsten in the divertor simulator NAGDIS-II. Helium plasma irradiation to tungsten led to the formation of nanostructures on the surface, while the nanostructures were annihilated after the potential of the specimen was changed to positive for several 10 min so that electrons irradiated the sample without ion irradiation. The specimens were analyzed in detail by transmission electron microscope with the help of focused ion beam technique. It is revealed that the helium nano-bubbles still remained even after the nanostructures were disappeared from the surface. Porosity of the nanostructured tungsten was measured from the TEM images.     

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.     

Lithium concentration dependence of implanted helium retention in lithium silicates

Helium ions of 500 keV were implanted with a fluence of 1.4 × 10¹⁷ ion/cm² into various lithium silicates to investigate whether a threshold level of helium retention exists in Li-containing silicate ceramics similar to that found in SiO_x in previous work. The composition and phases of the as prepared lithium silicates were determined by proton backscattering spectrometry (p-BS) and X-ray diffraction (XRD) methods with an average error of ±10%. Electrostatic charging of the samples was successfully eliminated by wrapping the samples in Al foil. The amounts of the retained helium within the samples were determined by subtracting the non-implanted spectra from the implanted ones. The experimental results show a threshold in helium retention depending on the Li concentration. Under 20 at.% all He is able to escape from the material; at around 30 at.% nearly half of the He, while over 65 at.% all implanted He is retained. With compositions expressed in SiO₂ volume percentages, a trend similar to those reported of SiO_x previously is found.