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 a...
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doaj-6d31a8704e184fdc991ce734f3efea512020-11-25T01:37:20ZengElsevierNuclear Materials and Energy2352-17912019-08-0120Retention properties in displacement damaged ultra-fine grain tungsten exposed to divertor plasmaJ.L. Barton0D.A. Buchenauer1W.R. Wampler2D.L. Rudakov3Z.Z. Fang4C.J. Lasnier5J.A. Whaley6J.G. Watkins7E.A. Unterberg8R.D. Kolasinski9H.Y. Guo10Sandia National Laboratories, P.O. Box 969, Livermore, CA 94551, USASandia National Laboratories, P.O. Box 969, Livermore, CA 94551, USASandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185, USAUniversity of California San Diego, 9500 Gilman Drive, La Jolla, CA 92093, USAUniversity of Utah, 201 President's Circle, Salt Lake City, UT 84112, USALawrence Livermore National Laboratory, 7000 East Avenue, Livermore, CA 94550, USASandia National Laboratories, P.O. Box 969, Livermore, CA 94551, USASandia National Laboratories, P.O. Box 969, Livermore, CA 94551, USAOak Ridge National Laboratory, 1 Bethel Valley Rd, Oak Ridge, TN 37830, USASandia National Laboratories, P.O. Box 969, Livermore, CA 94551, USA; Corresponding author.General Atomics, P.O. Box 85608, San Diego, CA 92186, USAOne 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 (<500 °C), providing an avenue for minimizing tritium retention in this material that would be practical for implementation in a tokamak.http://www.sciencedirect.com/science/article/pii/S2352179118301704 |
collection |
DOAJ |
language |
English |
format |
Article |
sources |
DOAJ |
author |
J.L. Barton D.A. Buchenauer W.R. Wampler D.L. Rudakov Z.Z. Fang C.J. Lasnier J.A. Whaley J.G. Watkins E.A. Unterberg R.D. Kolasinski H.Y. Guo |
spellingShingle |
J.L. Barton D.A. Buchenauer W.R. Wampler D.L. Rudakov Z.Z. Fang C.J. Lasnier J.A. Whaley J.G. Watkins E.A. Unterberg R.D. Kolasinski H.Y. Guo Retention properties in displacement damaged ultra-fine grain tungsten exposed to divertor plasma Nuclear Materials and Energy |
author_facet |
J.L. Barton D.A. Buchenauer W.R. Wampler D.L. Rudakov Z.Z. Fang C.J. Lasnier J.A. Whaley J.G. Watkins E.A. Unterberg R.D. Kolasinski H.Y. Guo |
author_sort |
J.L. Barton |
title |
Retention properties in displacement damaged ultra-fine grain tungsten exposed to divertor plasma |
title_short |
Retention properties in displacement damaged ultra-fine grain tungsten exposed to divertor plasma |
title_full |
Retention properties in displacement damaged ultra-fine grain tungsten exposed to divertor plasma |
title_fullStr |
Retention properties in displacement damaged ultra-fine grain tungsten exposed to divertor plasma |
title_full_unstemmed |
Retention properties in displacement damaged ultra-fine grain tungsten exposed to divertor plasma |
title_sort |
retention properties in displacement damaged ultra-fine grain tungsten exposed to divertor plasma |
publisher |
Elsevier |
series |
Nuclear Materials and Energy |
issn |
2352-1791 |
publishDate |
2019-08-01 |
description |
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 (<500 °C), providing an avenue for minimizing tritium retention in this material that would be practical for implementation in a tokamak. |
url |
http://www.sciencedirect.com/science/article/pii/S2352179118301704 |
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