Beryllium as a Plasma Facing Material for Near-Term Fusion Devices

Federici, G.; Barabash, V.; Doerner, R. P.; Lorenzetto, P.; Matthews, G. F.; Raffray, A.R.

doi:10.1016/b978-0-12-803581-8.09805-2

Cited by 7 publications

(6 citation statements)

References 214 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…However, despite its plasma compatible properties, Be has an increased sputtering yield when subjected to intense D and T particle bombardment and this will consequently lead to the erosion, migration and re-deposition of Be layers. Co-deposition of Be together with nuclear fuel underlines the T retention problem as co-deposited layers will exhibit different retention and release properties [14][15][16]. Furthermore, it is expected that Be co-deposited layers will be the main contributor to T retention in ITER [17,18].…”

Section: Introductionmentioning

confidence: 99%

Deuterium Retention and Release Behavior from Beryllium Co-Deposited Layers at Distinct Ar/D Ratio

et al. 2021

View full text Add to dashboard Cite

Beryllium-deuterium co-deposited layers were obtained using DC magnetron sputtering technique by varying the Ar/D2 gas mixture composition (10/1; 5/1; 2/1 and 1:1) at a constant deposition rate of 0.06 nm/s, 343 K substrate temperature and 2 Pa gas pressure. The surface morphology of the layers was analyzed using Scanning Electron Microscopy and the layer crystalline structure was analyzed by X-ray diffraction. Rutherford backscattering spectrometry was employed to determine the chemical composition of the layers. D trapping states and inventory quantification were performed using thermal desorption spectroscopy. The morphology of the layers is not influenced by the Ar/D2 gas mixture composition but by the substrate type and roughness. The increase of the D2 content during the deposition leads to the deposition of Be-D amorphous layers and also reduces the layer thickness by decreasing the sputtering yield due to the poisoning of the Be target. The D retention in the layers is dominated by the D trapping in low activation binding states and the increase of D2 flow during deposition leads to a significant build-up of deuterium in these states. Increase of deuterium flow during deposition consequently leads to an increase of D retention in the beryllium layers up to 300%. The resulted Be-D layers release the majority of their D (above 99.99%) at temperatures lower than 700 K.

show abstract

Section: Introductionmentioning

confidence: 99%

Deuterium Retention and Release Behavior from Beryllium Co-Deposited Layers at Distinct Ar/D Ratio

et al. 2021

View full text Add to dashboard Cite

show abstract

“…Beryllium's advantages as a plasma facing material are its low Z, good thermal conductivity, and high oxygen gettering characteristics. It has been tested as a plasma facing material also in earlier experiments both in JET, with its first introduction in 1989, as well as a smaller early period tokamaks such as UNITOR and ISX-B [2]. The main drawback of the use of beryllium is its high toxicity.…”

Section: Plasma Facing Components In Jet Iter Like Wallmentioning

confidence: 99%

“…By storing these samples in vacuum for 40 days 16% of tritium is desorbed from carbon samples (mostly in the form of DT and T2). Tritium is mostly accumulated by physical trapping in W samples and after 18 days and 250 days 50% and up to 70% of the accumulated tritium had diffused out of the W samples, respectively [2].…”

Section: Chemical State Of Tritium Accumulated In Bementioning

confidence: 99%

Tritium in plasma-facing components of JET with the ITER-Like-Wall

et al. 2021

View full text Add to dashboard Cite

The ITER-Like-Wall project has been carried out at the Joint European Torus (JET) to test plasma facing materials relevant to ITER. Materials being tested include both bulk metals (Be and W) and coatings. Tritium accumulation mechanisms and release properties depend both on the wall components, their location in the vacuum vessel, conditions of exposure to plasma and to the material itself. In this study, bulk beryllium limiter tiles, plasma-facing beryllium coated Inconel components from the main chamber, bulk tungsten and tungsten coated carbon fibre composite divertor tiles were analysed. A range of methods have been developed and applied in order to obtain a comprehensive overview on tritium retention and behaviour in different materials of plasma facing components (PFCs). Tritium content and chemical state were studied by the means of chemical or electrochemical dissolution methods and thermal desorption spectroscopy. Tritium distribution in the vacuum vessel and factors affecting its accumulation have been assessed and discussed.

show abstract

“…Beryllium choice is based on its low Z, good thermal conductivity, and high oxygen gettering characteristics. Beryllium has been tested as a plasma facing material in the currently largest tokamak device -Joint European Torus JET (first introduction in 1989, since 2012 -ITER-Like-Wall (ILW) project [8]) and also in smaller early period tokamaks such as UNITOR and ISX-B [9].…”

Section: Introductionmentioning

confidence: 99%

“…gettering. Be has been tested as a PFC material in the largest current tokamak-the Joint European Torus (JET) [9] (first introduced in 1989 and since 2012 in the ITER-like-wall (ILW) project [10]) and also in smaller early period tokamaks such as UNITOR and ISX-B [11].…”

Section: Introductionmentioning

confidence: 99%

Novel method for determination of tritium depth profiles in metallic samples

et al. 2019

View full text Add to dashboard Cite

Novel method for determining the depth profile of tritium in metallic samples has been demonstrated. Tritium accumulation in the fusion reactor materials is considered as a radiological issue due to its radioactivity. Therefore, tritium behavior prediction and estimation of its overall retention in fusion devices is of high importance. Proposed method in this study allows to measure depth profile of tritium in the metallic samples after exposure to tritium containing plasma, tritium gas or after irradiation with neutrons resulting in the tritium formation. In the method, successive layers of metal are removed using appropriate etching agent in the controlled regime and amount of evolved gases measured by the means of chromatography (gas composition and release rate) and proportional gas flow detector (tritium). Results on tritium profile in neutron irradiated, plasma exposed and tritium gas loaded beryllium have been reported and possible applications of the method for other metallic samples have been tested within this research.

show abstract

Beryllium as a Plasma Facing Material for Near-Term Fusion Devices

Cited by 7 publications

References 214 publications

Deuterium Retention and Release Behavior from Beryllium Co-Deposited Layers at Distinct Ar/D Ratio

Deuterium Retention and Release Behavior from Beryllium Co-Deposited Layers at Distinct Ar/D Ratio

Tritium in plasma-facing components of JET with the ITER-Like-Wall

Novel method for determination of tritium depth profiles in metallic samples

Contact Info

Product

Resources

About