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0.8 MeV and 6 MeV iron (Fe) ions were implanted into tungsten (W) to produce the irradiation damages with the various damage distributions. Thereafter, 1.0 keV deuterium ion (D2+) implantation was performed to evaluate the D retention behavior on damage distribution in W. The experimental results showed that the total D retentions were decreased by increasing the damage concentration introduced near the surface region by 0.8 MeV Fe ion implantation. The retention of D trapped by vacancy clusters and voids, which would be the stable trapping sites with higher trapping energies, were reduced, suggesting that the recombination of D atom into D2 on the W surface was enhanced due to D accumulation near the surface region. It can be said that the hydrogen retention behavior in PFMs will be controlled by the damage distribution near the surface.

Tungsten (W) samples were damaged by neutron and 6.4 MeV Fe-ion irradiation above 1000 K simulating the divertor operation temperature. Deuterium (D) retention properties were examined by decorating the damaged W with D and subsequent thermal desorption spectroscopy (TDS) measurements. Vacancy clusters were the major D trapping site in the W irradiated with Fe-ion at 873 K, although D retention by vacancy clusters decreased in the W irradiated with Fe-ion at 1173 K due to dynamic annealing. The D de-trapping activation energy from vacancy clusters was found to be 1.85 eV. D retention in neutron damage W was larger than that damaged by Fe-ion due to the uniform distribution of irradiation defects. The D desorption behaviors from neutron damaged W was simulated well by assuming the D de-trapping activation energy to be 1.52 eV.

Deuterium (D) retention behaviors for the 3 keV Helium (He+) implanted damaged-Tungsten (W) and 10 keV Carbon (C+) - 3 keV He+ simultaneous implanted damaged-W were evaluated by thermal desorption spectroscopy (TDS) to understand the synergetic effect of defect formation and C/He existence on D retention behavior for W with various damage level. For the He+ implantation, the retention of D trapped by dislocation loops was controlled by 3 keV He+ fluence. The D retention in the deeper region was reduced by He+ implantation with higher He+ fluence due to the formation of He bubbles and dense defects at the surface region which would reduce the effective D diffusion coefficient. In addition, in the case of the simultaneous C+ - He+ implantation, the reduction of D retention trapped in the deeper region was also found by the higher C+ - He+ fluence. It can be said that D retention behavior was controlled by the formation of He induced defects and accumulation of He near the surface even if the damages were introduced in the deeper region.

Additive Manufacturing (AM) can create novel and complex engineered material structures. Features such as controlled porosity, micro-fibers and/or nano-particles, transitions in materials and integral robust coatings can be important in developing solutions for fusion subcomponents. A realistic understanding of this capability would be particularly valuable in identifying development paths. Major concerns for using AM processes with lasers or electron beams that melt powder to make refractory parts are the power required and residual stresses arising in fabrication. A related issue is the required combination of lasers or e-beams to continue heating of deposited material (to reduce stresses) and to deposit new material at a reasonable built rate while providing adequate surface finish and resolution for meso-scale features. Some Direct Write processes that can make suitable preforms and be cured to an acceptable density may offer another approach for PFCs.

Six single crystal tungsten specimens were neutron irradiated to a dose of 0.1 displacements per atom (dpa) at three different irradiation temperatures (633 K, 963 K, and 1073 K) at the High Flux Isotope Reactor in Oak Ridge National Laboratory under the US-Japan PHENIX project. A pair of neutron-irradiated tungsten specimens was exposed to deuterium (D) plasma to D ion fluence of 5.0 × 1025 m−2 at three different exposure temperatures (673 K, 873 K, and 973 K) at the Tritium Plasma Experiment in Idaho National Laboratory. A combination of thermal desorption spectroscopy, nuclear reaction analysis, and rate-diffusion modeling code (Tritium Migration Analysis Program, TMAP) were used to understand D behavior in neutron-irradiated tungsten. A broad D desorption spectrum from the plasma-exposure temperature up to 1173 K was observed. Total D retention up to 1.9 × 1021 m−2 and near-surface D concentrations up to 1.7 × 10−3 D/W were experimentally measured from the 0.1 dpa neutron-irradiated single crystal tungsten. Trap density up to 2.0 × 10−3 Trap/W and detrapping energy ranging from 1.80 to 2.60 eV were obtained from the TMAP modeling.

To evaluate Helium (He) effect on hydrogen isotope retention in tungsten (W), He+ was introduced into W bulk by 201 – 1000 keV He+, or W surface by 3 keV He+ for 6 MeV Fe ion damaged W at room temperature. The deuterium (D) retention behavior was evaluated by thermal desorption spectroscopy (TDS). In addition, the amount of tritium (T) at surface and bulk were separately evaluated by beta-ray induced X-ray spectroscopy (BIXS). The experimental results indicated that the formation of He-void complexes reduced the D trapping in vacancies and voids which have higher trapping energy by the bulk He retention. The BIXS measurement also supported the He enhanced the D reduction in the W bulk region. On the other hand, the He ion irradiation near the surface region enhanced D trapping by dislocation loops or surface, indicating the existence of He near surface interfered the D diffusion toward the bulk. It was concluded that the He existence in bulk or surface will significantly change the D trapping and diffusion behavior in damaged W. Keywords: He in damaged W, Hydrogen isotope retention behavior, Damaged W, Fusion

An overview of recent Plasma-Material Interactions (PMI) research at the DIII-D tokamak using the Divertor Material Evaluation System (DiMES) is presented. The DiMES manipulator allows for exposure of material samples in the lower divertor of DIII-D under well-diagnosed ITER-relevant plasma conditions. Plasma parameters during the exposures are characterized by an extensive diagnostic suite including a number of spectroscopic diagnostics, Langmuir probes, IR imaging, and Divertor Thomson Scattering. Post-mortem measurements of net erosion/deposition on the samples are done by Ion Beam Analysis, and results are modelled by the ERO and REDEP/WBC codes with plasma background reproduced by OEDGE/DIVIMP modelling based on experimental inputs. This article highlights experiments studying sputtering erosion, re-deposition and migration of high-Z elements, mostly tungsten and molybdenum, as well as some alternative materials. Results are generally encouraging for use of high-Z PFCs in ITER and beyond, showing high redeposition and reduced net sputter erosion. Two methods of high-Z PFC surface erosion control, with (i) external electrical biasing and (ii) local gas injection, are also discussed. These techniques may find applications in the future devices.

Deuterium (D) retention behavior for the sequential 6 MeV iron (Fe) and 10 keV carbon (C) implanted tungsten (W) were evaluated by thermal desorption spectroscopy (TDS) and β-ray-induced X-ray spectroscopy (BIXS) to understand the synergetic effect of defect formation and C existence on D retention behavior for W under various damage distribution profiles. The experimental results indicated that retention of D trapped by dislocation loops was controlled by 10 keV C + implantation. The D retention was reduced in the sequential Fe 2+ − C + implanted W with higher C + fluence in comparison to that with lower C + fluence due to the formation of C-W layer which suppressed D diffusion toward the bulk and dense defects at the surface which reduce effective D diffusion coefficient. On the other hand, the amount of D trapped by the defects in the deeper region than C + implantation region (50 nm) was increased due to the formation of dense defects by 6 MeV Fe 2+ implantation within the depth of 1.5 μm.

The annealing effects on deuterium (D) retention for 0.1–1.0 dpa iron (Fe) ion damaged W were studied as a function of annealing duration. The D2 spectra for Fe damaged W with lower defect concentr...

We present measurements and modeling of aluminum erosion and redeposition experiments in separate helium and deuterium low power, low density L-mode plasmas at the outer divertor strike point of DIII-D to provide a low-Z material benchmark dataset for tokamak erosion-deposition modeling codes. Coatings of Al ∼100nm thick were applied to ideal (smooth) and realistic (rough) surfaces and exposed to repeat plasma discharges using the DiMES probe. Redeposition in all cases was primarily in the downstream toroidal field direction, evident from both in-situ spectroscopic and post-mortem non-spectroscopic measurements. The gross Al erosion yield was estimated from film thickness change measurements of small area samples, and was found to be ∼40–70% of the expected erosion yield based on theoretical physical sputtering yields after including sputtering by a 1–3% carbon impurity. The multi-step redeposition and re-erosion process, and hence the measured net erosion yield and material migration patterns, were found to be influenced by the surface roughness and/or porosity. A time-dependent model of material migration accounting for deposit accumulation in hidden areas was developed to reproduce the measurements in these experiments and determine a redeposition probability distribution function for sputtered atoms. Keywords: Erosion, Deposition, Migration, Roughness, Aluminum, Mixed materials

Future tokamak devices are envisioned to utilize a high-Z metal divertor with tungsten as the leading candidate. However, tokamak experiments with tungsten divertors have seen significant detriment...

The PHENIX Project is a 6-year U.S./Japan bilateral, multi-institutional collaboration program for the Technological Assessment of Plasma Facing Components for DEMO Reactors. The goal is to address the technical feasibility of helium-cooled divertor concepts using tungsten as the armor material in fusion power reactors. The project specifically attempts to (1) improve heat transfer modeling for helium-cooled divertor systems through experiments including steady-state and pulsed high-heat-load testing, (2) understand the thermomechanical properties of tungsten metals and alloys under divertor-relevant neutron irradiation conditions, and (3) determine the behavior of tritium in tungsten materials through high-flux plasma exposure experiments. The High Flux Isotope Reactor and the Plasma Arc Lamp facility at Oak Ridge National Laboratory, the Tritium Plasma Experiment facility at Idaho National Laboratory, and the helium loop at Georgia Institute of Technology are utilized for evaluation of the respo...

A compact electron cyclotron resonance plasma source has been utilized at Sandia National Laboratory to expose heated W samples (1270 K) to 50–75 eV He ions at fluxes on the order of 1019 m−2 s−1 a...

In this study, we measured deuterium (D) gas-driven permeation through tungsten (W) foils that had been pre-damaged by helium ions (He + ). The goal of this work was to determine how ion-induced damage affects hydrogen isotope permeation. At 873 K, the D permeability for W irradiated by 3.0 keV He + was approximately one order of magnitude lower than that for un-damaged W. This difference diminished with increasing temperature. Even after heating to 1173 K, the permeability returned to less than half of the value measured for un-damaged W. We propose that this is due to nucleation of He bubbles near the surface which potentially serve as a barrier to diffusion deeper into the bulk. Exposure at higher temperatures shows that the D permeability and diffusion coefficients return to levels observed for undamaged material. It is possible that these effects are linked to annealing of defects introduced by ion damage, and whether the defects are stabilized by the presence of trapped He.

This paper reviews recent results pertaining to hydrogen isotope behavior in neutron and heavy ion damaged W. Accumulation of damage in W creates stable trapping sites for hydrogen isotopes, thereby changing the observed desorption behavior. In particular, the desorption temperature shifts higher as the defect concentration increases. In addition, the distribution of defects throughout the sample also changes the shape of TDS spectrum. Even if low energy traps were distributed in the bulk region, the D diffusion toward the surface requires additional time for trapping/detrapping during surface-to-bulk transport, contributing to a shift of desorption peaks toward higher temperatures. It can be said that both of distribution of damage (e.g. hydrogen isotope trapping sites) and their stabilities would have a large impact on desorption. In addition, transmutation effects should be also considered for an actual fusion environment. Experimental results show that production of Re by nuclear reaction of W with neutrons reduces the density of trapping sites, though no remarkable retention enhancement is observed.

Effects of annealing after/under iron (Fe) ion irradiation on deuterium (D) retention behavior in tungsten (W) were studied. The D2 TDS spectra as a function of heating temperature for 0.1 dpa damaged W showed that the D retention was clearly decreased as the annealing temperature was increased. In particular, the desorption of D trapped by voids was largely reduced by annealing at 1173 K. The TEM observation indicated that the size of dislocation loops was clearly grown, and its density was decreased by the annealing above 573 K. After annealing at 1173 K, almost all the dislocation loops were recovered. The results of positron annihilation spectroscopy suggested that the density of vacancy-type defects such as voids, was decreased as the annealing temperature was increased, while its size was increased, indicating that the D retention was reduced by the recovery of the voids. Furthermore, it was found that the desorption temperature of D trapped by the voids for damaged W above 0.3 dpa was shifted toward higher temperature side. These results lead to a conclusion that the D retention behavior is controlled by defect density. The D retention in the samples annealed during irradiation was less than that annealed after irradiation. This result shows that defects would be quickly annihilated before stabilization by annealing during irradiation.

The Tritium Plasma Experiment (TPE) is a unique high-flux linear plasma device that can handle beryllium, tritium, and neutron-irradiated plasma facing materials, and is the only existing device dedicated to evaluate in-vessel tritium inventory in the nuclear environment for fusion safety. The electrical upgrade were recently carried out to enhance operational safety and to improve plasma performance. New DC power supplies and a new control center enable remote plasma operations from outside of the contamination area for tritium, minimizing the possible exposure risk with tritium and beryllium and eliminating heat stress issue. In November 2015, the TPE successfully achieved first deuterium plasma via remote operation after a significant three-year upgrade. Simple linear scaling estimate showed that the TPE is expected to achieve Γimax of >1.0 × 1023 m−2 s−1 and qheat of >1 MW m−2 with new power supplies. This upgrade not only improves operational safety of the worker, but also enhances plasma performance to better simulate extreme plasma-material conditions expected in ITER, FNSF, and DEMO for improving in-vessel tritium inventory assessment in fusion nuclear environment.

In this work, we examine the response of an ultra-fine grained (UFG) tungsten material to high-flux deuterium plasma exposure. UFG tungsten has received considerable interest as a possible plasma-facing material in magnetic confinement fusion devices, in large part because of its improved resistance to neutron damage. However, optimization of the material in this manner may lead to trade-offs in other properties. We address two aspects of the problem in this work: (a) how high-flux plasmas modify the structure of the exposed surface, and (b) how hydrogen isotopes become trapped within the material. The specific UFG tungsten considered here contains 100 nm-width Ti dispersoids (1 wt%) that limit the growth of the W grains to a median size of 960 nm. Metal impurities (Fe, Cr) as well as O were identified within the dispersoids; these species were absent from the W matrix. To simulate relevant particle bombardment conditions, we exposed specimens of the W-Ti material to low energy (100 eV), high-flux (> 1022 m− 2 s− 1) deuterium plasmas in the PISCES-A facility at the University of California, San Diego. To explore different temperature-dependent trapping mechanisms, we considered a range of exposure temperatures between 200 °C and 500 °C. For comparison, we also exposed reference specimens of conventional powder metallurgy warm-rolled and ITER-grade tungsten at 300 °C. Post-mortem focused ion beam profiling and atomic force microscopy of the UFG tungsten revealed no evidence of near-surface bubbles containing high pressure D2 gas, a common surface degradation mechanism associated with plasma exposure. Thermal desorption spectrometry indicated moderately higher trapping of D in the material compared with the reference specimens, though still within the spread of values for different tungsten grades found in the literature database. For the criteria considered here, these results do not indicate any significant obstacles to the potential use of UFG tungsten as a plasma-facing material, although further experimental work is needed to assess material response to transient events and high plasma fluence.

To address the transport and trapping of hydrogen isotopes, several permeation experiments are being pursued at both Sandia National Laboratories (deuterium gas-driven permeation) and Idaho National Laboratories (tritium gas- and plasma-driven tritium permeation). These experiments are in part a collaboration between the US and Japan to study the performance of tungsten at divertor relevant temperatures (PHENIX). Here we report on the development of a high temperature (≤1150 °C) gas-driven permeation cell and initial measurements of deuterium permeation in several types of tungsten: high purity tungsten foil, ITER-grade tungsten (grains oriented through the membrane), and dispersoid-strengthened ultra-fine grain (UFG) tungsten being developed in the US. Experiments were performed at 500–1000 °C and 0.1–1.0 atm D2 pressure. Permeation through ITER-grade tungsten was similar to earlier W experiments by Frauenfelder (1968–69) and Zaharakov (1973). Data from the UFG alloy indicates marginally higher permeability (

An associated particle neutron generator is described that employs a negative ion source to produce high neutron flux from a small source size. Negative ions produced in an rf-driven plasma source are extracted through a small aperture to form a beam which bombards a positively biased, high voltage target electrode . Electrons co-extracted with the negative ions are removed by a permanent magnet electron filter. The use of negative ions enables high neutron output (100% atomic ion beam), high quality imaging (small neutron source size), and reliable operation (no high voltage breakdowns). The neutron generator can operate in either pulsed or continuous-wave (cw) mode and has been demonstrated to produce 10 6 D–D n/s (equivalent to ~108 D–T n/s) from a 1 mm-diameter neutron source size to facilitate high fidelity associated particle imaging.

Aluminum (Al) is a convenient proxy for beryllium (Be) plasma material interaction studies since they have a number of physical and chemical similarities. Al samples were exposed at the lower outer strike point of an L-mode divertor plasma in DIII-D (conditions 7–11 × 1018 D-ions cm−2 s−1, Te = 12–47 eV). The gross erosion rate was directly measured using post-mortem ion beam analysis of small 1 mm-sized samples where local re-deposition was determined to be negligible. The gross erosion rate was also calculated using spectroscopic methods, but these rates greatly underestimate the direct (i.e. non-spectroscopic) measurement. The direct measured erosion yields were within the range of published D+ → Al ion beam sputtering yields. The ionizations per photon (S/XB) coefficients used in the spectroscopic analysis were determined in separate experiments using He plasmas at the PISCES-B linear plasma facility at UCSD. The measured S/XB coefficients were on average ∼6× higher than the theoretically calculated values.

In this work, we apply low energy ion beam analysis to examine directly how the adsorbed hydrogen concentration and binding configuration on W(1 0 0) depend on temperature. We exposed the tungsten surface to fluxes of both atomic and molecular H and D. We then probed the H isotopes adsorbed along different crystal directions using 1–2 keV Ne+ ions. At saturation coverage, H occupies two-fold bridge sites on W(1 0 0) at 25 °C. The H coverage dramatically changes the behavior of channeled ions, as does reconstruction of the surface W atoms. For the exposure conditions examined here, we find that surface sites remain populated with H until the surface temperature reaches 200 °C. After this point, we observe H rapidly desorbing until only a residual concentration remains at 450 °C. Development of an efficient atomistic model that accurately reproduces the experimental ion energy spectra and azimuthal variation of recoiled H is underway.

This paper reviews recent achievement of hydrogen isotope behavior for damaged tungsten. To demonstrate neutron irradiation, the irradiation damages were introduced into W by energetic Fe2+ irradiation and D retention behavior was examined by thermal desorption spectroscopy (TDS). The D trapping behavior was evaluated using Hydrogen Isotope Diffusion and Trapping (HIDT) code. It was found that D trapping states consisted of two-four stages with the trapping energy of 0.60 eV, 0.85 eV, 1.15-1.25 eV and 1.55 eV depending on the damage concentration and distribution. Based on these experimental results, the hydrogen isotope retention behavior in actual fusion condition was demonstrated. It was found that most of hydrogen isotope was retained in tungsten wall even if the wall temperature was kept at operation temperature.

Additively manufactured (AM) austenitic stainless steels are intriguing candidates for the storage of gaseous hydrogen isotopes because complex vessel geometries can be built more easily than by using conventional machining options. Parts built with AM stainless steel tend to have excellent mechanical properties (with tensile strength, ductility, fatigue crack growth, and fracture toughness comparable to or exceeding that of wrought austenitic stainless steel). However, the solidification microstructures produced by AM processing differ substantially from the microstructures of wrought material. Some features may affect permeability, including some amount of porosity and a greater amount of ferrite. Because the diffusivity of hydrogen in ferrite is greater than in austenite (six orders of magnitude at ambient temperature), care must be taken to retain the performance that is taken for granted due to the base alloy chemistry. Furthermore, AM parts tend to have greater dislocation densities and greater amounts of carbon, nitrogen, and oxygen. These features, along with the austenite/ferrite interfaces, may contribute to greater hydrogen trapping. We report the results of our studies of deuterium transport in various austenitic (304L, 316, and 316L) steels produced by AM. Manufacturing by Powder Bed Fusion (PBF) and two different blown powder methods are considered here (Laser Engineered Net Shaping® (LENS®) and a Direct Laser Powder Deposition (DLPD) method with a higher laser power)). The hydrogen permeability (an equilibrium property) changes negligibly (less than a factor of 2), regardless of chemistry and processing method, when tested between 150 and 500 °C. This is despite increases in ferrite content up to FN = 2.7. However, AM materials exhibit greater hydrogen isotope trapping, as measured by permeation transients, thermal desorption spectra, and inert gas fusion measurement. The trapping energies are likely modest (

Experiments on retention of hydrogen isotopes (including tritium) at temperatures less than 800 °C have been carried out in the Tritium Plasma Experiment (TPE) at Idaho National Laboratory [1] , [2] . To provide a direct measurement of plasma driven permeation in plasma facing materials at temperatures reaching 1000 °C, a new TPE membrane holder has been built to hold test specimens (≤1 mm in thickness) at high temperature while measuring tritium permeating through the membrane from the plasma facing side. This measurement is accomplished by employing a carrier gas that transports the permeating tritium from the backside of the membrane to ion chambers giving a direct measurement of the plasma driven tritium permeation rate. Isolation of the membrane cooling and sweep gases from TPE's vacuum chamber has been demonstrated by sealing tests performed up to 1000 °C of a membrane holder design that provides easy change out of membrane specimens between tests. Simulations of the helium carrier gas which transports tritium to the ion chamber indicate a very small pressure drop (∼700 Pa) with good flow uniformity (at 1000 sccm). Thermal transport simulations indicate that temperatures up to 1000 °C are expected at the highest TPE fluxes.

One of the main advantages of using tungsten (W) as a plasma facing material (PFM) is its low uptake and retention of tritium. However, in high purity (ITER grade) W, hydrogenic retention increases significantly with neutron-induced displacement damage in the W lattice. This experiment examines an alternative W grade PFM, ultra-fine grain (UFG) W, to compare its retention properties with ITER grade W after 12 MeV Si ion displacement damage up to 0.6 dpa (displacements per atom.) Following exposure to plasma in the DIII-D divertor, D retention was then assessed with Nuclear Reaction Analysis (NRA) depth profiling up to 3.5 µm and thermal desorption spectrometry (TDS). Undamaged specimens were also included in our test matrix for comparison. For all samples, D release peaks were observed during TDS at approximately 200 °C and 750 °C. For the ITER-grade W specimens, the intensity of the 750 °C release peak was more pronounced for specimens that had been pre-damaged. Conversely, UFG samples that had been damaged by 12 MeV Si showed enhancement of the lower temperature release peak (200 °C). NRA profiles also reveal a higher D concentration for UFG W samples up to the peak in the damage profile at a depth of 2 μm. Overall, we observed that the total trapped inventory in UFG W was 20% higher than ITER grade W in the undamaged case and 10% higher in the damaged case. A comparison of NRA and TDS data indicates that a larger fraction of the total retained D is trapped near the surface (86–100%) in UFG W pre-damaged to 0.6 dpa compared with ITER grade W (39–61%). Further examination of the UFG material with microscopy is recommended for a definitive determination of the types of defects responsible for D trapping. Our results highlight some potential trade-offs associated UFG W regarding its performance from a tritium retention standpoint. That said, our TDS results indicate that this enhanced inventory can be released by baking at relatively low temperatures (

Comparisons have been made between heat flux measurements from Langmuir probes and embedded thermocouples in the divertor of DIII-D. Good agreement has been found near the outer strike point (OSP) during L-mode operation with Neutral Beam Injection (NBI) using a sheath power transmission factor (SPTF) of 7, predicted by collisionless 1-D sheath theory. Previous SPTF measurements taken from Langmuir probes and IR imagery on DIII-D demonstrated values below the theoretical limit. The Langmuir probe array has since been upgraded and an embedded thermocouple array has been utilized to measure heat flux. The SPTF has also been measured during a NBI heated H-mode shot. This shot demonstrated a SPTF greater than 7 neat the OSP, which is due to a larger scrape-off layer (SOL) current density during H-mode operation. These studies represent a significant advancement towards finding agreement between theoretical predictions of the SPTF at the divertor and experimental measurements from the divertor diagnostics.

Experimental observation of net erosion of molybdenum being significantly reduced compared to gross erosion in the divertor of DIII-D is reported for well-controlled plasma conditions. For the first time, gross erosion rates were measured by both spectroscopic and non-spectroscopic methods. In one experiment a net erosion rate of 0.73 ± 0.03 nm/s was measured using ion beam analysis (IBA) of a 1 cm diameter Mo-coated sample. For a 1 mm diameter Mo sample exposed at the same time the net erosion rate was higher at 1.31 nm/s. For the small sample redeposition is expected to be negligible in comparison with the larger sample yielding a net to gross erosion estimate of 0.56 ± 12%. The gross rate was also measured spectroscopically (386 nm MoI line) giving 2.45 nm/s ± factor 2. The experiment was modeled with the REDEP/WBC erosion/redeposition code package coupled to the ITMC–DYN mixed-material code, with plasma conditions supplied by the OEDGE code using Langmuir probe data input. The code-calculated net/gross ratio is =0.46, in good agreement with experiment.

The net erosion of molybdenum by the divertor plasma in the DIII-D tokamak was determined from the reduction in thickness of a thin film test sample after a short exposure to well controlled plasma conditions. The spatial distribution of Mo deposited on adjacent carbon surfaces was also measured. Integration of the total quantity of Mo deposited within 2 cm of the source, gave only 19% of the amount lost from the film indicating that most of the Mo is transported to greater distances, in spite of the short pathlength for ionization of Mo in the divertor plasma. These measurements provide benchmark data for comparisons between gross and net erosion and between measurements and simulations of erosion and deposition, which are discussed in companion papers at this conference. Erosion and deposition of carbon, and deuterium retention were also examined.

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.

Beryllium deuteride has been measured in plasma-exposed surfaces and in the plasma column of the PISCES-B experiment. XPS measurements of high-purity BeH2 samples, used as standards, confirm the presence of beryllium deuteride in beryllium codeposited material (Be 1s energy of BeH2 bond = 115.0 eV). TDS measurements of high-purity BeD2 samples are used to accurately determine the decomposition temperature of the beryllium deuteride molecule (∼540 K). Deuterium release is observed from plasma-exposed beryllium samples and from Be codeposited layers at this temperature. BeD molecular (A–X) band emission is observed in beryllium seeded plasma discharges and during deuterium sputtering measurements of beryllium. Band emission intensities during Be-seeding are used to estimate the photon emission coefficient, , for the band. The value verifies that molecular sputtering of BeD (from beryllium targets saturated with deuterium from the plasma exposure) is a small fraction of the physical sputtering of beryllium atoms.

In this study, the PISCES-A linear plasma instrument has been used to characterize retention in several carbon fiber composites in order to better understand the factors which lead to elevated retention levels in these materials. The PISCES instrument is capable of subjecting materials to intense fluxes (up to 10 22 m −2 s −1 ) of low energy (150 eV) D + ions, producing conditions similar to those encountered by plasma facing components in a fusion reactor. In this investigation, three CFCs (fabricated with different manufacturing processes) are compared with the N11 composite used in the Tore Supra reactor. The specific surface areas for these materials were within the range of 0.14–0.55 m 2 /g. The plasma bombardment conditions were adjusted to provide doses on the order of 10 25 –10 26 m −2 at a sample temperature of 200 °C. After removal from PISCES-A, the amount of D retained in the sample surface was determined via thermal desorption spectroscopy. The measured retention showed a strong correlation with the type of material used and the corresponding BET surface area. By using a CFC with a lower internal porosity, one could expect a reduction in retention by a factor of 5 or more.

Erosion of the main chamber plasma-facing components is of concern for ITER. Plasma interaction with the outboard chamber wall is studied in DIII-D using Langmuir probes and optical diagnostics. Fast camera data shows that edge localized modes (ELMs) feature helical filamentary structures propagating towards the outboard wall. Upon reaching the wall, filaments result in regions of local intense plasma–material interaction (PMI) where peak incident particle and heat fluxes are up to two orders of magnitude higher than those between ELMs. In low density/collisionality H-mode discharges, PMI at the outboard wall is almost entirely due to ELMs. A moderate change of the gap between the separatrix and the outer wall strongly affects PMI intensity at the wall. Material samples exposed near the outboard wall showed net carbon deposition in high-density discharges (near the Greenwald limit) and tendency towards net erosion in lower density discharges (∼0.45 of the Greenwald limit).

The Compact Toroid Injection Experiment (CTIX) produces a high density, high velocity hydrogen plasma that maintains its configuration in free space on a MHD resistive time scale. In order to study the production and acceleration of impurities in the injector, several sets of silicon collector probes were exposed to spheromak-like CT’s exiting the accelerator. Elemental analysis by Auger Electron Spectroscopy indicated the presence of O, Al, Fe, and Cu in films up to 200 A thickness (1000 CT interactions). Using a smaller number of CT interactions (10–20), implantation of Fe and Cu was measured by Auger depth profiling. The amount of impurities was found to increase with accelerating voltage and number of CT interactions while use of a solenoidal field reduced the amount. Comparison of the implanted Fe and Cu with TRIM simulations indicated that the impurities were traveling more slowly than the hydrogen CT.

An understanding of the behavior of hydrogen isotopes in materials is critical to predicting tritium transport in structural metals (at high pressure), estimating tritium losses during production (fission environment), and predicting in-vessel inventory for future fusion devices (plasma driven permeation). Current models often assume equilibrium diffusivity and solubility for a class of materials (e.g. stainless steels or aluminum alloys), neglecting trapping effects or, at best, considering a single population of trapping sites. Permeation and trapping studies of the particular castings and forgings enable greater confidence and reduced margins in the models. For FY15, we have continued our investigation of the role of ferrite in permeation for steels of interest to GTS, through measurements of the duplex steel 2507. We also initiated an investigation of the permeability in work hardened materials, to follow up on earlier observations of unusual permeability in a particular region of 304L forgings. Samples were prepared and characterized for ferrite content and coated with palladium to prevent oxidation. Issues with the poor reproducibility of measurements at low permeability were overcome, although the techniques in use are tedious. Funding through TPBAR and GTS were secured for a research grade quadrupole mass spectrometer (QMS) and replacement turbo pumps, whichmore » should improve the fidelity and throughput of measurements in FY16.« less

The heat fluxes expected in the ITER divertor raise concerns about the damage tolerances of tungsten, especially due to thermal transients caused by edge localized modes (ELMs) as well as frequent temperature cycling from high to low extremes. Therefore we are motivated to understand the heat flux conditions that can cause not only enhanced erosion but also bulk thermo-mechanical damage to a tungsten divertor. For the metal ring campaign in DIII-D, tungsten-coated TZM tile inserts were installed making two toroidal arrays of metal tile inserts in the lower divertor. This study examines the deposited heat flux on these rings with embedded thermocouples (TCs) sampling at 10 kHz and compares them to Langmuir probe (LP) and infrared thermography (IRTV) heat flux measurements. We see agreement of the TC, LP, and IRTV data within 20% of the heat flux averaged over the entire discharge, and that all three diagnostics suggest parallel heat flux at the OSP location increases linearly with input heating power. The TC and LP heat flux time traces during the discharge trend together during large changes to the average heat flux. By subtracting the LP measured inter-ELM heat flux from TC data, using a rectangular ELM energy pulse shape, and taking the relative size and duration of each ELM from measurements, we extract the ELM heat fluxes from TC data. This over-estimates the IRTV measured ELM heat fluxes by a factor of 1.9, and could be due to the simplicity of the TC heat flux model and the assumed ELM energy pulse shape. ELM heat fluxes deposited on the inserts are used to model tungsten erosion in this campaign. These TC ELM heat flux estimates are used in addition to IRTV, especially in cases where the IRTV view to the metal ring is obstructed. We observe that some metal inserts were deformed due to exposed leading edges. The thermal conditions on these inserts are investigated with the thermal modeling code ABAQUS using our heat flux measurements when these edges were exposed. We discuss how the thermal cycling on the ends of the inserts caused this deformation.

In this paper we report on density and impurity measurements in the Sustained Spheromak Physics Experiment (SSPX) which has recently started operation. The SSPX spheromak plasma is sustained by coaxial helicity injection for a duration of 2msec with peak toroidal currents of up to 0.5MA. The plasma-facing components consist of tungsten-coated copper to minimize sputtering. The surfaces are conditioned by a combination of baking at 150 C, glow discharge cleaning, Titanium gettering, and pulse-discharge cleaning with helium plasmas. In this way we can achieve density control so that the plasma density ({approx} 1-4 x 10{sup 20}m{sup -3}) matches the gas input. Low-density operation is presently limited by breakdown requirements, but we hope that new gas valves with supersonic nozzles will allow for a further reduction in density. We find that the conditioning reduces the impurity radiation to the point where it is no longer important to the energy balance, and long-lived spheromak plasmas are obtained (decay times of 1.5msec).

The Sustained Spheromak Physics Experiment (SSPX) will examine the confinement properties of spheromak plasmas sustained by DC helicity injection. Understanding the plasma-surface interactions is an important component of the experimental program since the spheromak plasma is in close contact with a stabilizing wall (flux conserver) and is maintained by a high current discharge in the coaxial injector region. Peak electron temperatures in the range of 400 eV are expected, so the copper plasma facing surfaces in SSPX have been coated with tungsten to minimize sputtering and plasma contamination. Here, we report on the characterization and conditioning of these surfaces used for the initial studies of spheromak formation in SSPX. The high pressure plasma-sprayed tungsten facing the SSPX plasma was characterized in situ using β-backscattering and ex situ using laboratory measurements on similarly prepared samples. Measurements showed that water can be desorbed effectively through baking while the removal rates of volatile impurity gases during glow discharge and shot conditioning indicated a large source of carbon and oxygen in the porous coating.

In this article we present an overview of the plasma diagnostics operating or planned for the sustained spheromak physics experiment device now operating at Lawrence Livermore National Laboratory. A set of 46 wall-mounted magnetic probes provide the essential data necessary for magnetic reconstruction of the Taylor relaxed state. Rogowski coils measure currents induced in the flux conserver. A CO2 laser interferometer is used to measure electron line density. Spectroscopic measurements include an absolutely-calibrated spectrometer recording extended domain spectrometer for obtaining time-integrated visible ultraviolet spectra and two time-resolved vacuum monochrometers for studying the time evolution of two separate emission lines. Another time-integrated spectrometer records spectra in the visible range. Filtered silicon photodiode bolometers provide total power measurements, and a 16 channel photodiode spatial array gives radial emission profiles. Two-dimensional imaging of the plasma and helicity injec...

In a magnetically diverted tokamak, the scrape-off layer (SOL) and divertor plasma separates the first wall from the core plasma, intercepting impurities generated at the wall before they reach the core plasma. The divertor plasma can also serve to spread the heat and particle flux over a large area of divertor structure wall using impurity radiation and neutral charge exchange, thus reducing peak heat and particle fluxes at the divertor strike plate. Such a reduction will be required in the next generation of tokamaks, for without it the divertor engineering requirements are very demanding. To successfully demonstrate a radiative divertor, a highly radiative condition with significant volume recombination must be achieved in the divertor, while maintaining a low impurity content in the core plasma. Divertor plasma properties are determined by a complex interaction of classical parallel transport, anomalous perpendicular transport, impurity transport and radiation, and plasma - wall interaction. In this paper we will describe a set of experiments on DIII-D designed to provide detailed two-dimensional documentation of the divertor and SOL plasma. Measurements have been made in operating modes where the plasma is attached to the divertor strike plate and in highly radiating cases where the plasma is detached from the divertor strike plate. We will also discuss the results of experiments designed to influence the distribution of impurities in the plasma using enhanced SOL plasma flow. Extensive modelling efforts will be described which are successfully reproducing attached plasma conditions and are helping to elucidate the important plasma and atomic physics involved in the detachment process.

Recent measurements of the two-dimensional (2-D) spatial profiles of divertor plasma density, temperature, and emissivity in the DIII-D tokamak [J. Luxon et al., in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1987), p. 159] under highly radiating conditions are presented. Data are obtained using a divertor Thomson scattering system and other diagnostics optimized for measuring the high electron densities and low temperatures in these detached divertor plasmas (ne⩽1021 m−3, 0.5 eV⩽Te). D2 gas injection in the divertor increases the plasma radiation and lowers Te to less than 2 eV in most of the divertor volume. Modeling shows that this temperature is low enough to allow ion–neutral collisions, charge exchange, and volume recombination to play significant roles in reducing the plasma pressure along the magnetic separatrix by a factor of 3–5, consistent with the measurements. Absolutely calibrated vacuum ultravi...

Deuterium fuelled radiative diverters, operating in the partially detached divertor (PDD) regime, are effective in reducing the heat load on the divertor target. These PDD discharges are shown to be compatible with active particle exhaust conditions-an important issue for ITER type operation. The measured global and local plasma properties of actively pumped PDD discharges are also shown to have important similarities with non-pumped PDD discharges. Active particle exhaust leads to quasi-equilibrium plasma conditions in both the divertor and the main body within several particle confinement times (i.e. ~800 ms). However, changes is the gas puffing and/or pumping programmes have an almost immediate impact on the divertor plasma (e.g., the PDD regime ends ~50 ms after the injected deuterium gas how is interrupted). In contrast to the pumping results with lower density non-radiative divertor discharges, the divertor cryopump particle exhaust rate is relatively insensitive to the location of the divertor separatrix strike point during radiative divertor operation. Edge localized modes (ELMs) are shown to trigger multifaceted asymmetric radiation from the edge (MARFE) at the divertor under slightly submarginal PDD conditions

Electron density, temperature and parallel pressure measurements at several locations along field lines connecting the midplane scrapeoff layer (SOL) with the outer divertor are presented for both attached and partially-detached divertor cases: Ip = 1.4 MA, q95 = 4.2 and Pinput ∼ 6.7 MW under ELMing H-mode conditions. At the onset of the Partially Detached Divertor (PDD), a high density, low temperature plasma forms in the divertor SOL (divertor MARFE). The electron pressure drops by a factor of 2 between the midplane separatrix and the X-point and then an additional ∼ 3–5 times between the X-point and the outboard separatrix strike point. These results are in contrast to the attached (non-PDD) case, where electron pressure in the SOL is reduced by, at most, a factor of two between the midplane and the divertor target. Divertor MARFEs generally have only marginal adverse impact on important H-mode characteristics, such as confinement time. In fact, PDD discharges at low input power (i.e., approximately twice the L-H-mode threshold power) maintain good H-mode characteristics until a high density, low temperature plasma abruptly forms inside the separatrix near the X-point (X-point MARFE). Concurrent with the appearance of this X-point MARFE is a degradation in both energy confinement and the plasma fueling rate and an increase in the carbon impurity concentration inside the core plasma. The formation of the X-point MARFE is consistent with a thermal instability resulting from the temperature dependence of the carbon radiative cooling rate in the range ∼ 7–30 eV.

This paper describes divertor density and temperature measurements using both a new reciprocating Langmuir probe (XPT-RCP) which plunges vertically above the divertor floor up to the X-point height and swept, single, Langmuir probes fixed horizontally across the divertor floor. These types of measurements are important for testing models of the SOL and divertor which then are used to determine engineering design criteria for plasma facing components in reactor size tokamaks. The 6 mm diameter fixed single probes (19 domed and 2 flat at radially equivalent locations) are incorporated into the lower divertor floor at 19 radial locations and swept at 250 Hz. These probes are critical for determining plasma detachment from the floor during operation with high density, highly radiating divertors. By sweeping the divertor strike point across the fixed probes, different regions of the target plate incident flux profile can be sampled and a high resolution spatial profile can be obtained from each probe tip as the strike point moves past. The X-point reciprocating probe (XPT-RCP) provides ne and Te profiles with high spatial (2 mm) and temporal (0.5 ms) resolution from the target plate to the X-point along a single vertical chord at the same radial location as a fixed probe tip at a different azimuthal location. The probe ne and Te are compared to the divertor Thomson scattering (DTS) ne and Te (eight vertical points at 20 Hz, RThomson = RX-point-rcp). Recent observations have also shown divertor densities from 3 × 1019 to 4 × 1020 m−3 near the target plate with the highest densities observed with D2 gas puffing. Electron temperature is typically of the order of 15–25 eV at the target rising to about 70 eV near the X-point. Lower temperature, higher density plasmas are observed along the inner leg. Generally good agreement among the XPT-RCP, the fixed floor probes, and the DTS is observed. Differences between these diagnostic measurements will also be discussed with respect to different operating regimes and conditions.

Net erosion rates at the outer strike point of the DIII-D divertor plasma were measured for several materials during quiescent H-mode operation with deuterium plasmas. Materials examined include graphite, beryllium, tungsten, vanadium and molybdenum. For graphite, net erosion rates up to 4 nm/sec were found. Erosion rates for the metals were much smaller than for carbon. Ion fluxes from Langmuir probe measurements were used to predict gross erosion by sputtering. Measured net erosion was much smaller than predicted gross erosion. Transport of metal atoms by the plasma across the divertor surface was also examined. Light atoms were transported farther than heavy atoms as predicted by impurity transport models.

A comprehensive set of experiments has been carried out on the DIII-D tokamak to measure the 2-D (R,Z) structure of the divertor plasma in a systematic way using new diagnostics. Measurements cover the divertor radially from inside the X-point to the outer target plate and vertically from the target plate to above the X-point. Identical, repeatable shots were made, each having radial sweeps of the X-point and divertor strike points, to allow complete plasma and radiation profile measurements. Data have been obtained in ohmic, L-mode, ELMing H-mode, and reversed B{sub T} operation ({gradient}B drift away from the X-point). In addition, complete measurements were made of radiative divertor plasmas with a Partially Detached Divertor (PDD) induced by D{sub 2} injection and with a Radiating Mantle induced by Impurity injection (RMI) using neon and nitrogen. The data set includes first observations of the radial and poloidal profiles of the X-point, inner and outer leg plasmas in PDD and RMI radiative divertor operation. Preliminary data analysis shows that intrinsic impurities play a critical role in determining the SOL and divertor conditions.

DIII-D currently operates with a single- or double-null open divertor and graphite walls. Active particle control with a divertor cryopump has demonstrated density control, efficient helium exhaust, and reduction of the inventory of particles in the wall. Gas puffing of D{sub 2} and impurities has demonstrated reduction of the peak divertor beat flux by factors of 3--5 by radiation. A combination of active cryopumping and feedback-controlled D{sub 2} gas puffing has produced similar divertor heat flux reduction with density control. Experiments with neon puffing have shown that the radiation is equally-divided between a localized zone near the X-point and a mantle around the plasma core. The density in these experiments has also been controlled with cryopumping. These experimental results combined with modeling were used to develop the new Radiative Divertor for DIII-D. This is a double-null slot divertor with four cryopumps to provide particle control and neutral shielding for high-triangularity advanced tokamak discharges. UEDGE and DEGAS simulations, benchmarked to experimental data, have been used to optimize the design.