Beryllium Plasma-facing Components Research Articles

Investigations into penetration depth profiles of hydrogenic species in beryllium plasma-facing components via molecular dynamics simulations

During the operation of nuclear fusion reactors, plasma-facing components lining the reactor vessel are continually bombarded by plasma species. The penetration and subsequent trapping of these bombarding plasma ions has implications for component damage as well as in-vessel inventory. Accurately predicting the expected ion penetration depth profiles at a range of plasma ion and surface temperatures typical of fusion reactor operating conditions will inform the scrape-off layer design to limit particle radiation damage and tritium trapping in order to prolong the lifetime of the plasma-facing components and satisfy the DT fuel cycle requirements. By defining a statistical distribution for ion penetration depth and describing the evolution of its parameters across the fusion parameter space of interest, the expected ion deposition depth profiles can be calculated for any subset of ion and surface temperature ranges as needed. Molecular dynamics simulations were used to study the bombardment of beryllium lattices with surface temperatures of up to 1100 K by 5 eV–150 eV deuterium and tritium ions, and the resulting ion penetration depths were investigated. The distributions of two penetration depth quantities, considered from the perspectives of lattice damage and hydrogen retention are defined and their distribution parameter dependence on surface and ion temperature is identified. The expected positive correlation between penetration depth and ion temperature is observed, where the non-linear relationship between these quantities indicates the expected form of the velocity dependence of nuclear stopping power at low bombardment energies. Isotope effects on the distributions are also investigated, with results suggesting that heavier ions have comparably lower mobility within the sample and will generally accumulate closer to the surface. A short study on ion deposition rates is also performed; a non-linear increase of deposition rate with increasing bombarding ion energy has been observed, and evidence of a weak positive surface temperature correlation has been noted.

Impact of water ingress on deuterium release, oxidation, and dust generation in beryllium plasma-facing components

Beryllium samples from the JET ITER-like wall limiter tiles with either co-deposits or surface cracks caused by melt damage, were immersed into boiling water for 4 h 15 min to simulate and assess the impact of coolant water ingress into a tokamak on the state of Be components. Microscopy of the water-treated surfaces and the lack of residue in the water revealed that no thermomechanical damage (cracking or exfoliation) occurred to the samples during the exposure. Ion beam analysis showed no measurable release of deuterium from the samples. Combined ion beam analysis and Raman spectroscopy indicated only some degree of surface oxidation, but no thick oxide films were formed.

Data on erosion and hydrogen fuel retention in Beryllium plasma-facing materials

ITER will use beryllium as a plasma-facing material in the main chamber, covering a total surface area of about 620 m2. Given the importance of beryllium erosion and co-deposition for tritium retention in ITER, significant efforts have been made to understand the behaviour of beryllium under fusion-relevant conditions with high particle and heat loads. This paper provides a comprehensive report on the state of knowledge of beryllium behaviour under fusion-relevant conditions: the erosion mechanisms and their consequences, beryllium migration in JET, fuel retention and dust generation. The paper reviews basic laboratory studies, advanced computer simulations and experience from laboratory plasma experiments in linear simulators of plasma–wall interactions and in controlled fusion devices using beryllium plasma-facing components. A critical assessment of analytical methods and simulation codes used in beryllium studies is given. The overall objective is to review the existing set of data with a broad literature survey and to identify gaps and research needs to broaden the database for ITER.

Estimates of RF-induced erosion at antenna-connected beryllium plasma-facing components in JET

Radio-frequency (RF)-enhanced surface erosion of beryllium (Be) plasma-facing components is explored, for the first time, using the ERO code. The code is applied to measured, RF-enhanced edge Be line emission at JET Be outboard limiters, in the presence of high-power, ion cyclotron-resonance heating (ICRH) in L-mode discharges. In this first modelling study, the RF sheath effect from an ICRH antenna on a magnetically connected, limiter region is simulated by adding a constant potential to the local sheath, in an attempt to match measured increases in local Be I and Be II emission of factors of 2–3. It was found that such increases are readily simulated with added potentials in the range of 100–200 V, which is compatible with expected values for potentials arising from rectification of sheath voltage oscillations from ICRH antennas in the scrape-off layer plasma. Absolute erosion values are also estimated within the uncertainties in local plasma conditions.

Runaway electron beam generation and mitigation during disruptions at JET-ILW

Disruptions are a major operational concern for next generation tokamaks, including ITER. They may generate excessive heat loads on plasma facing components, large electromagnetic forces in the machine structures and several MA of multi-MeV runaway electrons. A more complete understanding of the runaway generation processes and methods to suppress them is necessary to ensure safe and reliable operation of future tokamaks. Runaway electrons were studied at JET-ILW showing that their generation dependencies (accelerating electric field, avalanche critical field, toroidal field, MHD fluctuations) are in agreement with current theories. In addition, vertical stability plays a key role in long runaway beam formation. Energies up to 20 MeV are observed. Mitigation of an incoming runaway electron beam triggered by massive argon injection was found to be feasible provided that the injection takes place early enough in the disruption process. However, suppressing an already accelerated runaway electron beam in the MA range was found to be difficult even with injections of more than 2 kPa.m3 high-Z gases such as krypton or xenon. This may be due to the presence of a cold background plasma weakly coupled to the runaway electron beam which prevents neutrals from penetrating in the electron beam core. Following unsuccessful mitigation attempts, runaway electron impacts on beryllium plasma-facing components were observed, showing localized melting with toroidal asymmetries.

Thermal behavior and temperature measurements of melting beryllium plasma-facing components exposed to high heat flux

The emissivity of metallic materials is low and varies with temperature and wavelength inducing errors on surface temperature measurements. High heat flux experiments on beryllium were carried out to investigate the thermal behavior of bulk Be tiles. Thermal modeling aiming at determining the surface and bulk temperatures have been performed using ANSYS®. A Be tile was exposed to heat flux with power density ranging between 1 and 7MW/m2. Surface temperatures were measured using an infrared camera in the 3–5μm range and two-color pyrometers, one at short wavelengths (1.5–1.7μm) and one at mid IR range wavelengths (2–4μm) range.Both the IR camera and two-color pyrometers do not provide accurate temperature measurements on melted Be due to changes in the emissivities and emissivity ratio induced by surface modifications.

Inter-machine comparison of the termination phase and energy conversion in tokamak disruptions with runaway current plateau formation and implications for ITER

The termination of the current and the loss of runaway electrons following runaway current plateau formation during disruptions have been investigated in the JET, DIII-D and FTU tokamaks. Substantial conversion of magnetic energy into runaway kinetic energy, up to ∼10 times the initial plateau runaway kinetic energy, has been inferred for the slowest current terminations. Both modelling and experiment suggest that, in present devices, the efficiency of conversion into runaway kinetic energy is determined to a great extent by the characteristic runaway loss time, τdiff, and the resistive time of the residual ohmic plasma after the disruption, τres, increasing with the ratio τdiff/τres. It is predicted that, in large future devices such as ITER, the generation of runaways by the avalanche mechanism will play an important role, particularly for slow runaway discharge terminations, increasing substantially the amount of energy deposited by the runaways onto the plasma-facing components by the conversion of magnetic energy of the runaway plasma into runaway kinetic energy. Estimates of the power fluxes on the beryllium plasma-facing components during runaway termination in ITER indicate that for runaway currents of up to 2 MA no melting of the components is expected. For larger runaway currents, minimization of the effects of runaway impact on the first wall requires a reduction in the kinetic energy of the runaway beam before termination and, in addition, high plasma density ne and low ohmic plasma resistance (long τres) to prevent large conversion of magnetic into runaway kinetic energy during slow current terminations.

Beryllium plasma-facing components for the ITER-like wall project at JET

ITER-Like Wall Project has been launched at the JET tokamak in order to study a tokamak operation with beryllium components on the main chamber wall and tungsten in the divertor. To perform this first comprehensive test of both materials in a thermonuclear fusion environment, a broad program has been undertaken to develop plasma-facing components and assess their performance under high power loads. The paper provides a concise report on scientific and technical issues in the development of a beryllium first wall at JET.

Beryllium R&D for fusion applications

Beryllium is one of the primary candidates as both plasma-facing material (PFM) and neutron multiplier in the next-step fusion reactors. Both sintered-product blocks and pebbles are considered in fusion reactor designs. Beryllium evaporated on carbon tiles has also been used in Joint European Torus (JET) and may be considered for other designs. Future efforts are directed toward the pebble form of beryllium. Research and evaluations of data are underway to determine the most attractive material processing approaches in terms of fabrication cost and quality; technical issues associated with heat transfer; thermal, mechanical and irradiation stability; safety and tritium release. Beryllium plasma-facing components will require periodic repair or replacement, therefore disposal or recycling of activated and tritiated beryllium will also be a concern. Beryllium as a component of the molten salt, Flibe is also being considered in novel approaches to the plasma–structure interface. This paper deals with the main issues related to the use of Be in a fusion reactor as both neutron multiplier and first wall material. These issues include potential reactions with steam during accidents and the health and environmental aspects of its use, reprocessing and reuse, or disposal.

Negative Transferred Arc Cleaning: A Method for Roughening and Removing Surface Contamination from Beryllium and Other Metallic Surfaces

TA cleaning has been investigated for preparing the surface of beryllium plasma facing components (PFC’s) inside of the International Thermonuclear Experimental Reactor (ITER) prior to depositing beryllium by plasma spraying. Plasma spraying of beryllium was evaluated during the ITER Engineering Design Activity (EDA) for in-situ repair and initial fabrication of the beryllium first wall armor. Results have shown that surface roughening of beryllium, during the TA cleaning process, can result in bond strengths greater than 100 MPa between beryllium surfaces and plasma sprayed beryllium. In addition, the TA cleaning process was shown to be an effective method for removing contaminate layers of carbon and tungsten from the surface of beryllium. Investigations have been performed to characterize the different arc-types that occur during the TA cleaning process (Type I, II and III arcs) and the effectiveness of the TA cleaning process for potentially removing co-deposited layers of carbon and deuterium from the surface of beryllium, stainless steel and tungsten.

Beryllium plasma-facing components: JET experience

JET has accumulated practical experience of plasma operation in both limiter and divertor configurations using massive beryllium plasma-facing components. These latter have shown evidence, even after a very short operational phase, for localised melting of the beryllium surface. Sustained melting results in substantial transport of the liquid beryllium metal; there is, however, no evidence for low cycle termal fatigue under this extreme environment. Design has proceeded by thermal testing of prototypes with analysis used for interpretation only. For flux densities ≤5 MW m −2, T surface ≤1000°C, typically 200 pulses of 6–8 s can be sustained with plastic deformation and no surface fatigue cracking. The latter is seen for more extreme conditions, i.e. flux densities in excess of 14 MW m −2, pulse lengths ≤0.5 s surface temperature >700°C. Failure of thin cladding of beryllium to heat sinks, which tends to occur along the beryllium to substrate interface, was insufficient to produce fatigue effects in the bulk beryllium material, at fluxes ≤10 MW m −2.