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Buchenauer, D. A.

Publications and source records attributed to Buchenauer, D. A..

3 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Localized divertor leakage measurements using isotopic tungsten sources during edge-localized mode-y H-mode discharges on DIII-D

Experiments carried out on DIII-D using a novel setup of isotopic tungsten (W) sources in the outer divertor have characterized how the W leakage from this region depends on both the exact source location and edge-localized mode (ELM) behavior. The sources are toroidally-symmetric and poloidally-localized to two regions: (1) the outer strike point (OSP) with natural abundance of W isotopes; and (2) the far-target with highly-enriched 182 W isotopes. With the use of a dual-faced collector probe (CP) in the main scrape-off layer (SOL) near the outside midplane and source-rate spectroscopy, a proxy for divertor impurity leakage is developed herein. Using this proxy, it is found that for the OSP W location, there is a nearly linear increase of leakage with the power across the separatrix ( ), which is consistent with the effect of an increased upstream ion temperature parallel gradient force in the near-SOL; trends in the pedestal density and collisionality are also seen. Conversely, it is found that for the far-target W location leakage falls off rapidly as increases and ELM size decreases, which is suggestive that ELM size plays a role in the leakage from this location. Indications for main SOL W contamination is evidenced by the measurement of large deposition asymmetries on the two opposite CP faces. These measurements are coupled with interpretive modeling showing SOL W accumulation near the separatrix furthest from both targets driven by forces parallel to the magnetic field. This experimental setup, together with the target and upstream W measurements, provides information on the transport from different divertor W source locations and leakage. These studies help to elucidate the physics driving divertor impurity source rates and leakage, with and without ELMs, and provide better insight on the link in the chain connecting wall impurity sources to core impurity levels in magnetic fusion devices.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2019

Retention properties in displacement damaged ultra-fine grain tungsten exposed to divertor plasma

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 m. 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 (<500 °C), providing an avenue for minimizing tritium retention in this material that would be practical for implementation in a tokamak.

Barton, J↗