Tungsten nanostructures (W-fuzz) prepared in the PISCES-A linear device have been found to survive direct exposure to divertor plasmas in DIII-D. W-fuzz was exposed in the lower divertor of DIII-D using the divertor material evaluation system. Two samples were exposed in lower single null (LSN) deuterium H-mode plasmas. The first sample was exposed in three discharges terminated by vertical displacement event disruptions, and the second in two discharges near the lowered X-point. More recently, three samples were exposed near the lower outer strike point in predominantly helium H-mode LSN plasmas. In all cases, the W-fuzz survived plasma exposure with little obvious damage except in the areas where unipolar arcing occurred. Arcing is effective in W-fuzz removal, and it appears that surfaces covered with W-fuzz can be more prone to arcing than smooth W surfaces.

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.

DiMES samples featuring 1 cm and 1 mm diameter W films deposited on a Si substrate were exposed in DIII-D near the attached outer strike point of lower single-null L-mode discharges. The measured net and gross erosion rates of W, determined from post-mortem ion beam analysis (IBA) of 1 cm and 1 mm samples, were 0.14 and 0.48 nm/s, respectively, giving net/gross erosion ratio of 0.29. REDEP/WBC modeling of this experiment yielded a very similar ratio of 0.33. In another experiment, Mo-coated samples were exposed with CH4 gas injected ~12 cm upstream of DiMES. Reduction of Mo erosion was evidenced in-situ by the suppression of MoI line radiation. Post-mortem IBA showed that the net erosion of Mo in the middle of the sample was below the measurement resolution of 0.5 nm, corresponding to a rate of ≤0.07 nm/s. Compared to the previously measured erosion rates, this constitutes a reduction of more than 10X. D2 gas injection in similar geometry resulted in MoI radiation reduction by a factor of 3-5, yet net erosion of Mo stayed relatively high.

An accelerated fusion energy development program, a "fast-track" approach, requires proceeding with a nuclear and materials testing program in parallel with research on burning plasmas, ITER. A Fusion Nuclear Science Facility (FNSF) would address many of the key issues that need to be addressed prior to DEMO, including breeding tritium and completing the fuel cycle, qualifying nuclear materials for high fluence, developing suitable materials for the plasma-boundary interface, and demonstrating power extraction. The Advanced Tokamak (AT) is a strong candidate for an FNSF as a consequence of its mature physics base, capability to address the key issues, and the direct relevance to an attractive target power plant. The standard aspect ratio provides space for a solenoid, assuring robust plasma current initiation, and for an inboard blanket, assuring robust tritium breeding ratio (TBR) >1 for FNSF tritium self-sufficiency and building of inventory needed to start up DEMO. An example design point gives a moderate sized Cu-coil device with R/a = 2.7 m/0.77 m, kappa = 2.3, B-T = 5.4T, I-p = 6.6 MA, beta(N) = 2.75 P-fus = 127 MW. The modest bootstrap fraction of f(BS) = 0.55 provides an opportunity to develop steady state with sufficient current drive for adequate control. Proceeding with a FNSF in parallel with ITER provides a strong basis to begin construction of DEMO upon the achievement of Q similar to 10 in ITER. (c) 2014 Elsevier B.V. All rights reserved.

A substantial reduction of net compared to gross erosion of molybdenum and tungsten was observed in experiments conducted in the lower divertor of DIII-D using the divertor material evaluation system. Post-exposure net erosion of molybdenum and tungsten films was measured by Rutherford backscattering (RBS) yielding net erosion rates of 0.4–0.7 nm s−1 for Mo and ∼0.14 nm s−1 for W. Gross erosion was estimated using RBS on a 1 mm diameter sample, where re-deposition is negligible. Net erosion on a 1 cm diameter sample was reduced compared to gross erosion by factors of ∼2 for Mo and ∼3 for W. The experiment was modeled with the REDEP/WBC erosion/re-deposition code package coupled to the Ion Transport in Materials and Compounds—DYNamics mixed-material code, with plasma conditions supplied by the Onion skin modeling + Eirene + Divimp for edGE modeling code with input from divertor Langmuir probes. The code-calculated net/gross erosion rate ratios of 0.46 for Mo and 0.33 for W are in agreement with the experiment.

A leading power reactor breeding blanket candidate for a fusion demonstration power plant (DEMO) being pursued by the US Fusion Community is the Dual Coolant Lead Lithium (DCLL) concept. The safety hazards associated with the DCLL concept as a reactor blanket have been examined in several US design studies. These studies identify the largest radiological hazards as those associated with the dust generation by plasma erosion of plasma blanket module first walls, oxidation of blanket structures at high temperature in air or steam, inventories of tritium bred in or permeating through the ferritic steel structures of the blanket module and blanket support systems, and the 210Po and 203Hg produced in the PbLi breeder/coolant. What these studies lack is the scrutiny associated with a licensing review of the DCLL concept. An insight into this process was gained during the US participation in the ITER Test Blanket Module (TBM) Program. In this paper we discuss the lessons learned during this activity and make safety proposals for the design of a Fusion Nuclear Science Facility (FNSF) or a DEMO that employs a lead lithium breeding blanket.

Experimental results are presented from the three-Langmuir probe (LP) diagnostic head of the divertor material evaluation system (DiMES) on DIII-D that confirm the size of the projected current collection area of the LPs, which is essential for properly measuring ion saturation current density (J(sat)) and the sheath power transmission factor (SPTF). Also using the 3-LP DiMES head, the hypothesis that collisional effects on plasma density occurring in the magnetic sheath of the tile are responsible for a lower than expected SPTF is tested and deemed not to have a significant impact on the SPTF. Three-dimensional thermal modeling of wall tiles is presented that accounts for lateral heat conduction, temperature dependence of tile material properties and radiative heat loss from the tile surface. This modeling was developed to be used in the analysis of temperature profiles of the divertor embedded thermocouple (TC) array to obtain more accurate interpretations of TC temperature profiles to infer divertor surface heat flux than have previously been accomplished using more basic one-dimensional methods.

Physics based integrated modelling of the baseline scenario for a Fusion Nuclear Science Facility based on the Advanced Tokamak concept (FNSF-AT) (Chan et al 2010 Fusion Sci. Technol. 57 66) has found steady-state equilibria with good stability and controllability properties at the fusion performance required to accomplish FNSF's nuclear science mission with margin. 2D divertor analysis for this baseline scenario predicts that peak heat flux <10 MW m(-2) can be obtained even with scrape-off layer power width similar to 1 mm. Using this baseline fusion performance, high fidelity and high-resolution 3D neutronics calculations show acceptable cumulative end-of-life organic insulator dose levels in all the device coils, and TBR > 1. Two current drive scenarios, two divertor configurations, and two blanket concepts have been analysed. FNSF-AT would complement ITER in addressing science and technology gaps to a commercially attractive DEMO, and could enable a DEMO construction decision triggered by the achievement of Q = 10 in ITER.

Reduced erosion of a high-Z PFC divertor surface was observed in DIII-D with local injection of methane and deuterium gases. Molybdenum-coated silicon samples were exposed in the lower divertor of DIII-D using DiMES under plasma conditions previously shown to cause significant net erosion of Mo. Three exposures with 13CH4 and one exposure with D2 gas injection about 12cm upstream of the samples located within 1–2cm of the attached strike point were performed. Reduction of Mo erosion was evidenced in-situ by the suppression of MoI line radiation at 386.4nm once the gas injection started. Post-mortem ion beam analysis demonstrated that the net erosion of molybdenum near the center of the samples exposed with 13CH4 injection was below the measurement resolution of 0.5nm, corresponding to a rate of ⩽0.04nm/s. Compared to the previously measured erosion rates, this constitutes a reduction by a factor of >10.

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 2cm 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.

Two types of arc tracks are observed on the plasma-facing components (PFCs) in DIII-D. “Unmagnetized” random walk tracks are produced during glow discharges; they are rare and have no importance for PFC erosion but may degrade diagnostic mirrors. “Magnetized” scratch-like type II tracks are produced by unipolar arcs during plasma operations; they are formed by “retrograde BxJ” motion of the cathode spot and are roughly perpendicular to the local magnetic field. Type II arcs cause measurable erosion of graphite, but based on the evidence available they are relatively small contributors to the total erosion of carbon in DIII-D compared to other mechanisms such as physical and chemical sputtering and ablation from leading edges. Erosion by arcing of tungsten films deposited on graphite samples was observed in Divertor Material Evaluation System (DiMES) experiments. New DiMES experiments aimed at time-resolved arc measurements are proposed.

Understanding of Plasma Surface Interactions (PSI) and the selection of suitable plasma facing materials are critical areas for current tokamak experiments and future D-T burning facilities including ITER and FNSF. In support of PSI studies, DIII-D uses the Divertor Materials Evaluation System (DiMES), which contains a removable probe where material samples can be exposed to as few as a single well-characterized plasma shot. Experiments, consisting of a carbon DiMES probe surface with metal coatings of Be, W, V, Mo or Al, have been exposed to the DIII-D lower divertor strike point plasma for cumulative discharge times of 4-20s. Extensive DIII-D divertor diagnostics provided well-characterized plasmas for modeling efforts. Experimental results were benchmarked with modeling codes to validate and extend the predictive capability of the codes. Reported in this paper are two recent experiments and results. The first is on the net and gross erosion of Mo coatings and the extension of these results to an extrapolated all Mo surface DIII-D machine. The second is on the exposure to vertical displacement discharges and X-point plasma discharges of W-fuzz buttons, which were prepared by the PISCES (UCSD) laboratory. The surprising results are the robustness of the W-fuzz and that W impurity was not detected in the plasma core at the conditions studied.

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±0.03nm/s was measured using ion beam analysis (IBA) of a 1cm diameter Mo-coated sample. For a 1mm diameter Mo sample exposed at the same time the net erosion rate was higher at 1.31nm/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 (386nm MoI line) giving 2.45nm/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.

Thermo-oxidation is controlled exposure in an oxygen-containing atmosphere at elevated temperature and is being considered as a technique for the detritiation of carbon-based codeposits in ITER. In addition, unplanned oxidation may also occur during accidental air ingress. The impact of thermo-oxidation on ITER diagnostic mirrors causes concerns. A dedicated study was performed in DIII-D, where molybdenum and copper mirrors were installed in the main chamber, in the divertor, and at a location remote from the plasma and exposed for similar to 2 hours to a mixture containing 80% helium and 20% oxygen at a total pressure of 1.27 kPa. Mirrors in the main chamber and in the divertor were exposed at 350 degrees C to 360 degrees C whereas the temperature of mirrors in the remote area was similar to 160 degrees C.Reflectivity of all mirrors was degraded after the thermo-oxidation showing a decrease in the UV range from 60% to 10% for molybdenum mirrors and a 90% drop for copper mirrors at the wavelength 250 nm. The reflectivity of mirrors exposed at lower temperature was less degraded. Surface analyses revealed formation of oxides on all mirrors.In ITER, shutters planned for mirror protection are ineffective against thermo-oxidation. Nevertheless, in-situ cleaning systems planned for ITER mirrors may efficiently remove oxide layers.

A Fusion Nuclear Science Facility (FNSF) is necessary to make possible a DEMO of the Advanced Tokamak (AT) type after ITER. One candidate, Fusion Nuclear Science Facility-AT (FNSF-AT), should have neutron wall loading of 1-2 MW/m(2), continuous operation for periods of up to two weeks, a duly factor goal of 0.3 on a year and neutron fluence of 3-6 MW-yr/m(2) in ten years to enable development of blankets suitable for tritium and electricity production while demonstrating nearly all the critical elements necessary for the qualification and design of a DEMO. FNSF-AT, also called FDF, will be designed using conservative implementations of all elements of AT physics to produce 150-300 MW fusion power with modest energy gain (Q<7) in a modest sized normal conducting coil device. It will demonstrate and its results will help in the selection of the DEMO tritium breeding blanket concept. It will demonstrate the tritium fuel cycle, the behavior of candidate plasma facing materials, and the design and cooling of the first wall chamber and divertor components. It will also provide experience in safe operation and remote maintenance necessary for the DEMO design.