Deuterium retention and thermal conductivity in ion-beam displacement-damaged tungsten

Tynan, G. R.; Doerner, R.; Barton, J.L.; Chen, Renkun; Cui, Shuang; Simmonds, M.; Wang, Yongqiang; Weaver, Jordan S.; Mara, Nathan A.; Pathak, Siddhartha

doi:10.1016/j.nme.2017.03.024

Cited by 21 publications

(10 citation statements)

References 38 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…Tynan et al . also indicated that pre-damaged tungsten increased deuterium retention 29 . Since the dislocations cannot reduce the hydrogen embrittlement in iron, the formation of voids during ion irradiation is a key point to improve the hydrogen embrittlement in the present study.…”

Section: Discussionmentioning

confidence: 88%

Novel Methods for Prevention of Hydrogen Embrittlement in Iron

2017

Sci Rep

View full text Add to dashboard Cite

Iron is the most widely used metal in the world. However, hydrogen embrittlement in steels—iron based alloys—is an important issue related to the safety of our infrastructure, such as railroads and bridges. Therefore, the prevention of hydrogen embrittlement in steels is necessary. In the present study, we demonstrate two novel methods for the prevention of hydrogen embrittlement in iron: one involves the low-energy implantation of helium, which is usually an element harmful to metals, into iron, the other is inducing damage to the iron surface by ion irradiation. In general, irradiation with high-energy particles leads to metal brittleness. In the former method, the driving force for hydrogen embrittlement in iron is weakened, in the latter method, hydrogen diffusion in iron is prevented because of trapping of hydrogen atoms in the vacancies produced by the irradiation. As a result, hydrogen embrittlement in iron was suppressed by both methods.

show abstract

Section: Discussionmentioning

confidence: 88%

Novel Methods for Prevention of Hydrogen Embrittlement in Iron

2017

Sci Rep

View full text Add to dashboard Cite

show abstract

“…Fu 4,a It is thought that this is caused by the so-called barrier effect in which the implanted He ions form a thin (few tens of nm) He nanobubble layer in the sample and act to reduce the D retention either via the formation of interconnected pathways to the sample surface, leading the D ions to diffuse back to the plasma and resulting in the D retention decreasing, or by the formation of a diffusion barrier to the D atoms, or both [6]. On the other hand, many studies [7][8][9][10][11] have shown that radiation damage in W induced e.g. using energetic heavy ions such as Cu and W or by intense neutron irradiation, followed by D 2 plasma exposure, results in a substantial increase in D retention relative to the retention in undamaged W. This is usually attributed to the fact that radiation damage induces new defects such as vacancies and dislocations, which supply additional traps for the diffusion D ions, resulting in the total D retention increasing.…”

Section: Nuclear Fusionmentioning

confidence: 99%

Reduced D trapping by plasma-implanted He nanobubbles in radiation damaged tungsten

Bai

Zheng

et al. 2019

Nucl. Fusion

Self Cite

View full text Add to dashboard Cite

The presence of a plasma-implanted helium nanobubble layer significantly reduces deuterium (D) retention in undamaged commercial ITER grade tungsten (W). In this paper, we show evidence that this phenomenon can survive displacement damage. A He plasma exposure (sample temperature 643 K, ion flux 10 22 m −2 s −1 at 100 eV, fluence 10 25 m −2 ) pre-treatment was performed to create a thin He nanobubble layer in the first ~15 nm of ITER grade W samples. Samples were then irradiated by 5 MeV Cu ions at room temperature to create 0.001 to 0.1 dpa with peak damage rates occurring about 860 nm below the sample surface. Samples without He plasma exposure pre-treatment were also irradiated by 5 MeV Cu ions to provide a controlled baseline. All samples were subsequently exposed to D plasma at 373 K to a fluence of 10 24 m −2 . Nuclear reaction analysis results show that across a range of peak dpa ranging from 0.001 to 0.1 dpa, D retention inventory in the samples with He plasma exposure pre-treatment is reduced by a factor of two compared to samples without He plasma exposure pre-treatment. Transmission electron microscopy directly showed a surviving nanobubble layer after 0.1 dpa damage. However, the thickness of the bubble layer appears to have been reduced. The results suggest that the plasma-implanted He nanobubble layer can survive radiation damage and still function to reduce D diffusion and retention in tungsten-based plasma facing components.

show abstract

“…TEM observations with similar proton irradiated materials showed increasing defect density with DPA up to a saturation limit around 0.2 DPA in tungsten from where on mostly the loop defect size increased with constant loop density [5]. Also [6] found saturation effects in the 0.2 DPA region with a factor 5.5 increase in retention at 300 K, but by using heavy ions with 1 μm damage range. Protons with ∼15 μm range showed a significantly higher factor of ∼13 increase for D retention in W [4].…”

Section: Introductionmentioning

confidence: 81%

Scaling of deuterium retention in <3 MeV proton damaged Beryllium, Eurofer, and W-5Re in the range of 0.0003 to 6 DPA

et al. 2021

View full text Add to dashboard Cite

In continuation of earlier work on 3 MeV proton-damaged tungsten and reduced-activation steels we present new results on Eurofer97, Beryllium and W-5%Re sintered alloy irradiated <400 K. Methodical improvements result in largely reduced uncertainties. Beryllium is loaded using a 5 kV D2 + ion-source to 6.3*1021 D m−2 at 300 K. Eurofer97 and W-5Re are loaded in PSI-2 to 3*1025 D m−2. Irradiation and D-loading are conducted at ∼400 K. The D retention is measured by 3He μ-NRA. An exponential saturation fits the W-5Re D-retention data with R2 = 0.99 . The retention increases by a factor 10.3 in W-5Re, similar as in W, but on a ~7 times lower level. Within ±25% uncertainty D Retention in Eurofer97 proves to be independent of displacement damage up to 6.3 DPA. Beryllium shows increased retention by a factor 3 up to the tested maximum of 0.08 DPA. The retention in beryllium starts saturating, but the limited DPA range allows fitting the data with exponentials and power-laws.

show abstract

Deuterium retention and thermal conductivity in ion-beam displacement-damaged tungsten

Cited by 21 publications

References 38 publications

Novel Methods for Prevention of Hydrogen Embrittlement in Iron

Novel Methods for Prevention of Hydrogen Embrittlement in Iron

Reduced D trapping by plasma-implanted He nanobubbles in radiation damaged tungsten

Scaling of deuterium retention in <3 MeV proton damaged Beryllium, Eurofer, and W-5Re in the range of 0.0003 to 6 DPA

Contact Info

Product

Resources

About