Temperature dependence of surface morphology and deuterium retention in polycrystalline ITER-grade tungsten exposed to low-energy,high-flux D plasma

Surface topography and deuterium retention in polycrystalline ITER-grade tungsten have been examined after exposure to a low-energy (38 eV/D), high-flux (10²² D/m² s) deuterium plasma with ion fluences of 10²⁶ and 10²⁷ D/m² at various temperatures. The methods used were scanning electron microscopy equipped with focused ion beam, thermal desorption spectroscopy, and the D(³He,p) ⁴He nuclear reaction at ³He energies varied from 0.69 to 4.0 MeV. During exposure to the D plasma at temperatures in the range from 320 to 815 K, small blisters of size in the range from 0.2 to 5 μm, depending on the exposure temperature and ion fluence, are formed on the W surface. At an ion fluence of 10²⁷ D/m², the deuterium retention increases with the exposure temperature, reaching its maximum value of about 10²² D/m² at 500 K, and then decreases below 10¹⁹ D/m² at 800 K.     

Influence of temperature and plasma composition on deuterium retention in refractory metals

Refractory materials are being considered potential candidates to build the first wall of the fusion reactor chamber. This work reports on the results of the study of tungsten and molybdenum metals exposed to high flux densities (～10²⁴ D/m² s) and low temperature (T_e ～ 3 eV) deuterium plasmas in Pilot-PSI irradiation facility.The hydrogenic retention in poly-crystalline W and Mo targets was studied with ³He nuclear reaction analyses (NRA). The NRA results clearly show a two-dimensional radial distribution of the deuterium with a minimum at the center and a maximum close to the edge. These distribution correlates well with the thermal profile of the sample surface, where a maximum of ～1600 K was measured at the center decreasing to ～1000 K in the edges. A maximum deuterium fluence retention of 5 × 10¹⁵ D/cm² was measured. The values of the retained fractions ranging from 10^?5 to 10^?6 D_retained/D_incident were measured with thermal desorption spectroscopy (TDS) and compares well with IBA results. Moreover, the presence of C in the plasma and its co-deposition increases the D retention in the region where a C film is formed. Both NRA and TDS results show no clear dependence of retention on incident fluence suggesting the absence of plasma related traps in W under these conditions.     

Hydrogen isotope exchange in tungsten: Discussion as removal method for tritium

Hydrogen isotope exchange in re-crystallized polycrystalline tungsten was investigated at 320 and 450 K. In a first step the tungsten samples were loaded with deuterium to a fluence of 10²⁴ D/m² from a low-temperature plasma at 200 eV/D particle energy. In a second step, H was implanted at the same particle energy and similar target temperature with a mass-separated ion beam at different ion fluences ranging from 2 × 10²⁰ to 7.5 × 10²³ H/m². The analytic methods used were nuclear reaction analysis with D(³He,p)α reaction and elastic recoil detection analysis with ⁴He. In order to determine the D concentration at depths of up to 7.4 μm the ³He energy was varied from 0.5 to 4.5 MeV. It was found that already at an H fluence of 2 × 10²⁰ H/m², i.e. at 1/5000 of the initial D fluence, about 30% of the retained D was released. Depth profiling of D without and with subsequent H implantation shows strong replacement close to the surface at 320 K, but extending to all analyzable depths at 450 K especially at high fluences, leading to higher release efficiency. The reverse sequence of hydrogen isotopes allowed the analysis of the replacing isotope and showed that the release of D is balanced by the uptake of H. It also shows that hydrogen does not diffuse through a region of filled traps into a region were unfilled traps can be encounter but transport is rather a dynamic process of trapping and de-trapping even at 320 K. Initial D retention in H loaded W is an order of magnitude higher than in pristine W, indicating that every H-containing trap is a potential trap for D. In consequence, hydrogen isotope exchange is not a viable method to significantly enhance the operation time before the tritium inventory limit is reached but should be considered an option to reduce the tritium inventory in ITER before major interventions at the end of an operation period.     

Carbon film growth and hydrogenic retention of tungsten exposed to carbon-seeded high density deuterium plasmas

Tungsten (W) targets have been exposed to high density (n_e ? 4 × 10¹⁹ m^?3), low temperature (T_e ? 3 eV) CH₄-seeded deuterium (D) plasma in Pilot-PSI. The surface temperature of the target was ～1220 K at the center and decreased radially to ～650 K at the edges. Carbon film growth was found to only occur in regions where there was a clear CII emission line, corresponding to regions in the plasma with T_e ? 2 eV. The maximum film thickness was ～2.1 μm after a plasma exposure time of 120 s. ³He nuclear reaction (NRA) analysis and thermal desorption spectroscopy (TDS) determine that the presence of a thin carbon film dominates the hydrogenic retention properties of the W substrate. Thermal desorption spectroscopy analysis shows retention increasing roughly linearly with incident plasma fluence. NRA measures a C/D ratio of ～0.002 in these films deposited at high surface temperatures.     

Effect of proton irradiation on deuterium diffusion in zirconium oxide layer

Deuterium diffusion in proton-irradiated oxide layer of zirconium alloy has been in situ examined at 573 K by using a combined technique of deuterium plasma exposure and nuclear reaction analysis of the D(³He,p)⁴He reaction. The zirconium alloy used was GNF-Ziron (a high iron Zircaloy-2 type alloy), which had been corroded in high temperature steam, producing an oxide layer of 1.6–1.7 μm thickness. Oxidized specimens were irradiated at ambient temperature with 350 keV H⁺ ions, and the total fluence was 1 × 10¹⁷ cm^?2. An outer non-protective oxide layer of 0.5–0.6 μm thickness, which was observed in the unirradiated oxide layer, evolved into the protective barrier oxide due to the proton irradiation. The evaluated diffusion coefficients in the barrier layer were almost identical for both the unirradiated and irradiated oxides. From X-ray diffraction measurements, lattice expansion and high compressive stress were found in the proton-irradiated oxide. The most probable mechanism for evolution of the diffusion property in the irradiated oxide was ascribed to the increase of the compressive stress induced by the constraint of the damage-accumulated oxide layer by the thick metal substrate.     

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.     

Nuclear micro-beam analysis of deuterium distribution in carbon fibre composites for controlled fusion devices

Probes made of carbon fibre composite NB41 were exposed to deuterium plasmas in the TEXTOR tokamak and in a simulator of plasma–wall interactions, PISCES. The aim was to assess the deuterium retention and its lateral and depth distribution. The analysis was performed by means of D(³He, p)⁴He and ¹²C(³He, p)¹⁴N nuclear reactions analysis using a standard (1 mm spot) and micro-beam (20 μm resolution). The measurements have revealed non uniform distribution of deuterium atoms in micro-regions: differences by a factor of 3 between the maximum and minimum deuterium concentrations. The differences were associated with the orientation and type of fibres for samples exposed in PICSES. For surface structure in the erosion zone of samples exposed to a tokamak plasma the micro-regions were more complex. Depth profiling has indicated migration of fuel into the bulk of materials.     

Investigating permeation and transport of H isotopes in tungsten by first-principles

We have investigated permeation and transport of hydrogen (H) isotopes in tungsten (W) single crystal employing first-principles calculations in junction with Fick’ law. Permeability was approximately evaluated according to the solubility and diffusion coefficient of H. The solubility for H in bulk W from present calculation is consistent with the experimental results measured by Frauenfelder. The permeation fluxes of H isotopes are examined at the different thickness of W wall. The permeation fluxes of deuterium with the W thickness of 21 μm at the temperature of 770 K and with the W thickness of 50 μm at the temperature of 893 K were 0.68 × 10¹³ atom/m²s and 0.34 × 10¹⁴ atom/m²s, respectively. The dissociation coefficients of H isotopes are also evaluated. We believe that the present first-principles combined with Fick’ law method can be also generalized to investigate permeation and transport of H isotopes in most metals since such H isotopes behaviors in most metals are similar to those of H isotopes in W.     

Damage evolution of yttria-stabilized zirconia induced by He irradiation

The study presents an investigation of damage evolution of yttria-stabilized zirconia (YSZ) induced by irradiation of 100 keV He ions at room temperature as a function of fluence. Transmission electron microscopy (TEM), X-ray diffraction (XRD) and atomic force microscopy (AFM) were used in order to study the nature and evolution of structural damage at different levels. Our study shows that various kinds of defects are formed with the increasing fluence. Firstly, at low fluences, from 1 × 10¹⁶ to 4 × 10¹⁶ cm^?2, of which maximum values of displacement per atom (dpa) range from 0.29 to 1.17, an elastic strain which is attributed to the accumulation of irradiation-induced discrete point defects, is presented. Secondly, in the intermediate fluences ranging from 8 × 10¹⁶ to 1 × 10¹⁷ cm^?2 with corresponding dpa varying from 2.33 to 2.91, a large drop of elastic strain occurs accompanied by presence of an intensive damage region, which is comprised by large and interacted defect clusters. Thirdly, at the two high fluences of 2 × 10¹⁷ and 4 × 10¹⁷ cm^?2, of which dpa are 5.83 and 11.65 respectively, a great amount of ribbon-like He bubbles with granular structure and cracks are presented at the depth of maximum concentration of deposited He atoms. The structural damage evolution and the mechanism of formation of He bubbles are discussed.     

Measurement of deuterium and helium by glow-discharge optical emission spectroscopy for plasma–surface interaction studies

To examine the resolution of isotope analysis of hydrogen with glow-discharge optical emission spectroscopy (GDOES), depth profiles of hydrogen and deuterium in a H-containing Ta/D-containing Ti/Ni layered structure were measured. The depth profiles of deuterium could be measured with sufficient resolution in the presence of relatively large amounts of hydrogen and vice versa. In addition, measurements of depth profiles of He implanted in W at room temperature were also performed with Ne plasma. The intensity of the He emissions was sufficiently high at a fluence of 10²⁰ He m^?2 or higher. The depth profiles of He measured in this manner were in good agreement with the results of cross-sectional observations using a transmission electron microscope. Therefore, it was concluded that GDOES with Ne plasma is a promising technique for the depth profile analysis of plasma-facing materials and deposited layers formed on them.     

Deuterium retention in tungsten exposed to low-energy, high-flux clean and carbon-seeded deuterium plasmas

Depth profiles of deuterium trapped in tungsten exposed to a low-energy (≈200 eV/D) and high deuterium ion flux (about 1 × 10²¹ D/m² s) in clean (We use the term ‘clean’ in quotation marks having in mind the impossibility to obtain absolutely clean plasma. In our case the conception ‘clean’ D plasma means the plasma without intentionally introduced carbon impurities.) and carbon-seeded D plasmas at an ion fluence of about 2 × 10²⁴ D/m² and various temperatures have been measured up to a depth of 7 μm using the D(³He, p)⁴He nuclear reaction at a ³He energy varied from 0.69 to 4.0 MeV. The deuterium retention in single-crystalline and polycrystalline W increases with the exposure temperature, reaching its maximum value at about 500 K (for ‘clean’ plasma) or about 600 K (for carbon-seeded plasma), and then decreases as the temperature grows further. It is assumed that tungsten carbide formed on the W surface under exposure to the carbon-seeded D plasmas serves as a barrier layer for diffusion and prevents the outward transport of deuterium, thus increasing the D retention in the bulk of tungsten.     

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].     

Fatigue lifetime of repaired high heat flux components for ITER divertor

Thermal fatigue behaviour of repaired monoblocks was assessed from High Heat Flux (HHF) tests up to 20 MW m^?2 on 11 components. Among these components, 8 monoblocks were repaired (2 CFC and 6 tungsten). These components were manufactured by two EU industries: ANSALDO Ricerche and PLANSEE. Non destructive examination was performed on SATIR thermography test bed before and after HHF tests. SATIR results show that repaired monoblocks have a good thermal exhaust capability before HHF tests. For all monoblocks, no degradation of thermal properties was noticed during cycles at 10 MW m^?2. After hundreds of cycles at 20 MW m^?2, two W repaired monoblock melted. Post-HHF SATIR examination revealed a degradation of thermal properties which is systematic for W melted monoblocks and non-systematic for W repaired ones. For CFC repaired monoblocks, no damage was observed up to 20 MW m^?2. For the first ITER divertor set, specifications for the pre-qualification are that CFC (Resp. W) components have to sustain in steady state 1000 cycles at 10 MW m^?2 (Resp. 3 MW m^?2) followed by 1000 cycles at 20 MW m^?2 (Resp. 5 MW m^?2). For the first ITER divertor set, the repair process is validated for CFC and W monoblocks.     

Influence of tungsten–carbon mixed layer and irradiation defects on deuterium retention behavior in tungsten

The D₂⁺ fluence dependence on deuterium (D) retention was studied to clarify the D retention mechanism in tungsten. The additional D desorption stage was observed around 660 K in the TDS spectrum for a sample implanted with D₂⁺ up to the fluence of 10²³ D⁺ m^?2, which desorption stage was not observed the D₂⁺ implanted sample with the fluence less than 10²² D⁺ m^?2. The TEM observation showed that the highly dense voids were formed in tungsten by D₂⁺ implantation with the fluence of 10²³ D⁺ m^?2, considering that the D would be trapped by voids. To understand the D trapping by voids in C⁺ implanted tungsten, C⁺–D₂⁺ sequential implantation experiments at various C⁺ implantation temperatures were performed. It was found that the amount of D desorbed around 560 K was increased by increasing the C⁺ implantation temperature. The formation of the voids was observed with increasing the C⁺ implantation temperature by TEM, indicating that the increase of D desorption around 560 K was caused by the formation of voids. However, the desorption temperature of D trapped by voids in C⁺ implanted sample was lower than that in D₂⁺ implanted one. TEM observation and XPS measurement indicated that this difference was caused by the increase of void size and/or the presence of implanted carbon.     

Annealing of ion implanted 4H–SiC in the temperature range of 100–800 °C analysed by ion beam techniques

Ion implantation induced damage formation and subsequent annealing in 4H–SiC in the temperature range of 100–800 °C has been investigated. Silicon Carbide was implanted at room temperature with 200 keV ⁴⁰Ar ions with two implantation fluences of 4 × 10¹⁴ and 2 × 10¹⁵ ions/cm². The samples were characterized by Rutherford backscattering and nuclear reaction analysis techniques in channeling mode using 2.00 and 4.30 MeV ⁴He ion beams for damage buildup and recovery in the Si and C sublattices, respectively. At low ion fluence, the restoration of the Si sublattice is evident already at 200 °C and a considerable annealing step occurs between 300 and 400 °C. Similar results have been obtained for the C sublattice using the nuclear resonance reaction for carbon, ¹²C(α,α)¹²C at 4.26 MeV. For samples implanted with the higher ion fluence, no significant recovery is observed at these temperatures.     

Dust particle formation due to interaction between graphite and helicon deuterium plasmas

The collection of dust particles using divertor simulation helicon plasmas has been carried out to examine dust formation due to the interaction between a graphite target and deuterium plasmas, which are planned to operate in the large helical device (LHD) at the Japanese National Institute for Fusion Science (NIFS). The collected dust particles are classified into three types: (i) small spherical particles below 400 nm in size, (ii) agglomerates whose primary particles have a size of about 10 nm, and (iii) large flakes above 1 μm in size. These features are quite similar to those obtained through hydrogen plasma operation, indicating that the dust formation mechanisms due to the interaction between a carbon wall and a plasma of deuterium, which is the isotope of hydrogen, is probably similar to those of hydrogen.     

High heat flux testing of mock-ups for a full tungsten ITER divertor

In order to evaluate the option to start the ITER operation with a full tungsten (W) divertor, the EU-DA launched an extensive R&D program. It consisted in its initial phase in the high heat flux (HHF) testing of W mock-ups and medium scale prototypes up to 20 MW/m² in the AREVA FE 200 facility (F). Critical heat flux (CHF) experiments were carried out on the items which survived the above thermal fatigue testing.After 1000 cycles at 10 MW/m², the full W Plasma Facing Components (PFCs) mock-ups successfully sustained either 1000 cycles at 15 MW/m² or 500 cycles at 20 MW/m².However, some significant surface melting, as well as the complete melting of a few monoblocks, occurred during the HHF thermal fatigue testing program representative of the present ITER requirements for the strike point region, namely 1000 cycles at 10 MW/m² followed by 1000 cycles at 20 MW/m².The results of the CHF experiments were also rather encouraging, since the tested items sustained heat fluxes in the range of 30 MW/m² in steady-state conditions.     

Development of W coatings for fusion applications

The paper gives a short overview on tungsten (W) coatings deposited by various methods on carbon materials (carbon fibre composite – CFC and fine grain graphite – FGG). Vacuum Plasma Spray (VPS), Chemical Vapor Deposition (CVD) and Physical Vapor Deposition (PVD) techniques are analyzed in respect with the characteristics and performances of the W coatings.A particular attention is paid to the Combined Magnetron Sputtering and Ion Implantation (CMSII) technique, which was developed during the last 4 years from laboratory to industrial scale and it is successfully applied for W coating (10–15 μm and 20–25 μm) of more than 2500 tiles for the ITER-like Wall project at JET and ASDEX Upgrade. This technique involves simultaneously magnetron sputtering and high energy (tens of keV) ion implantation. Due to the ion bombardment a stress relief occurs within the coating enabling its growth without delamination to a relatively large thickness. In addition, in order to adjust the thermal expansion mismatch between CFC and W, a Mo interlayer of 2–3 μm is currently used. Experimentally, W/Mo coatings with a thickness up to 50 μm were produced and successfully tested in the GLADIS ion beam facility up to 23 MW/m².     

Uptake of deuterium in partially D-depleted residual codeposits left on DIII-D tiles after laboratory thermo-oxidation experiments

Recent evidence has shown that tokamak carbon-based codeposits may become partially or fully depleted of hydrogen through thermo-oxidation, as the hydrogen content of the codeposits is removed more rapidly than the carbon content. In this study we examine the ability of such partially-depleted residual DIII-D divertor codeposits to uptake deuterium upon subsequent exposure to deuterium gas or deuterium plasmas. The partially D-depleted specimens used here were obtained from a previous study where DIII-D codeposits were oxidized for 2 h at 623 K (350 °C) and 267 Pa (2 Torr) O₂ [J.W. Davis et al., Thermo-oxidation of DIII-D codeposits on open surfaces and in simulated tile gaps, J. Nucl. Mater. 415 (2011) S789–S792]. In the present study some of these specimens, having undergone prior oxidation, were exposed to D₂ glow discharge plasmas or D₂ gas at 20 kPa (150 Torr) at 300 or 523 K. In the case of plasma exposure, no uptake of D was observed, while an increase in D content was seen following D₂ gas exposures. When the gas exposure took place at 300 K, heating the specimens in vacuum to 623 K for 15 min led to the release of all of the increased D content. For the gas exposure at 523 K, the increase in D content was found to require longer (8 h) vacuum baking to remove. However, in a reference codeposit specimen (from a closeby location on the tile), which had not been previously oxidized, there was a similar increase in D content following D₂ exposure at 523 K, but it could not be released even following 8 h vacuum baking at 623 K.     

Aerosol formation and hydrogen co-deposition by colliding ablation plasma plumes of lithium and lead

In a high-repetition inertial fusion reactor, along with pellet implosions, the interior of target chamber is to be exposed to high-energy, short pulses of X-ray, unburned DT and He ash particles and pellet debris. As a result, wall materials will be subjected to ablation, ejecting particles in the plasma state to collide with each other in the center of volume. The interaction dynamics of ablation plasmas of lithium and lead, candidate first wall materials, has been investigated in the deposited energy density range from 3 to 10 J/cm²/pulse at a repetition rate of 10 Hz, each 6 ns long. The plasma density and electron temperature of colliding ablation plumes have been found to vary from the order of 10⁸–10¹³ 1/cm³ and from 0.7 to 1.5 eV, respectively. The formation of aerosol in the form of droplet has been observed with diameters between 100 nm and 10 μm. Also, hydrogen co-deposition has been found to occur particularly for colliding plumes of lithium, resulting in the H/Li atomic ratio from 0.15 to 0.27 in the hydrogen partial pressure range from 10 to 50 Pa.