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

Conference · Thu Aug 01 00:00:00 EDT 2019 · Nuclear Materials and Energy

DOI:https://doi.org/10.1016/j.nme.2019.100689· OSTI ID:1561608

Barton, J ^[1]; Buchenauer, Dean ^[1]; Wampler, W. R. ^[1]; Rudakov, D.L. ^[2]; Fang, Zhigang ^[3]; Lasnier, C.J. ^[4]; Whaley, J A. ^[1]; Watkins, J. G. ^[1]; Unterberg, Ezekial  ^[5]; Kolasinski, R. D. ^[1]; Guo, H. Y. ^[6]

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1. Sandia National Laboratories (SNL)
2. University of California, San Diego
3. University of Utah
4. Lawrence Livermore National Laboratory (LLNL)
5. ORNL
6. General Atomics, San Diego

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.

• Research Organization: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
• Sponsoring Organization: USDOE Office of Science (SC)
• DOE Contract Number: AC05-00OR22725
• OSTI ID: 1561608
• Journal Information: Nuclear Materials and Energy, Vol. 20; Conference: 23rd International Conference on Plasma Surface Interactions in Controlled Fusion Devices (PSI-23) - Princeton, New Jersey, United States of America - 6/17/2018 4:00:00 AM-6/22/2018 4:00:00 AM, Princeton, New Jersey, United States of America , 6/17/2018 - 6/22/2018; ISSN 2352-1791
• Publisher: Elsevier
• Country of Publication: United States
• Language: English