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

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

doi:10.1016/j.nme.2019.100689

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

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

Power Exhaust and Particle Control

Research output: Contribution to journal › Article › peer-review

2 Scopus citations

Abstract

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.

Original language	English
Article number	100689
Journal	Nuclear Materials and Energy
Volume	20
DOIs	https://doi.org/10.1016/j.nme.2019.100689
State	Published - Aug 2019

Funding

Sandia is a multi-program laboratory managed and operated by Sandia Corporation, a wholly-owned subsidiary of the Honeywell Corporation, for the United States Department of Energy's National Nuclear Security Administration under contract DE-NA0003525 . This work was also supported by our coauthors under Department of Energy contracts DE-FG02-07ER54917 , DE-AC52-07NA27344 , DE-AC05-000R22725 , and DE-FC02-04ER54698 . Sandia is a multi-program laboratory managed and operated by Sandia Corporation, a wholly-owned subsidiary of the Honeywell Corporation, for the United States Department of Energy's National Nuclear Security Administration under contract DE-NA0003525. This work was also supported by our coauthors under Department of Energy contracts DE-FG02-07ER54917, DE-AC52-07NA27344, DE-AC05-000R22725, and DE-FC02-04ER54698.

Funders	Funder number
U.S. Department of Energy	DE-AC05-000R22725, DE-NA0003525, DE-FC02-04ER54698, DE-AC52-07NA27344, DE-FG02-07ER54917

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Barton, J. L., Buchenauer, D. A., Wampler, W. R., Rudakov, D. L., Fang, Z. Z., Lasnier, C. J., Whaley, J. A., Watkins, J. G., Unterberg, E. A., Kolasinski, R. D., & Guo, H. Y. (2019). Retention properties in displacement damaged ultra-fine grain tungsten exposed to divertor plasma. Nuclear Materials and Energy, 20, Article 100689. https://doi.org/10.1016/j.nme.2019.100689

@article{10547c8a878f46f890ab5922bde3db74,

title = "Retention properties in displacement damaged ultra-fine grain tungsten exposed to divertor plasma",

abstract = "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.",

author = "Barton, {J. L.} and Buchenauer, {D. A.} and Wampler, {W. R.} and Rudakov, {D. L.} and Fang, {Z. Z.} and Lasnier, {C. J.} and Whaley, {J. A.} and Watkins, {J. G.} and Unterberg, {E. A.} and Kolasinski, {R. D.} and Guo, {H. Y.}",

note = "Publisher Copyright: {\textcopyright} 2019 Elsevier Ltd",

year = "2019",

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doi = "10.1016/j.nme.2019.100689",

language = "English",

volume = "20",

journal = "Nuclear Materials and Energy",

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T1 - Retention properties in displacement damaged ultra-fine grain tungsten exposed to divertor plasma

AU - Barton, J. L.

AU - Buchenauer, D. A.

AU - Wampler, W. R.

AU - Rudakov, D. L.

AU - Fang, Z. Z.

AU - Lasnier, C. J.

AU - Whaley, J. A.

AU - Watkins, J. G.

AU - Unterberg, E. A.

AU - Kolasinski, R. D.

AU - Guo, H. Y.

PY - 2019/8

Y1 - 2019/8

N2 - 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.

AB - 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.

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

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