Beryllium Limiter Research Articles

Damage to JET beryllium tiles

JET has operated with beryllium limiters such that up to 180 MJ could be coupled to the plasma. Approximately 2–4% of the surface of the limiter has been melted near the plasma contact point. Another 10–15% of the surface shows evidence of edge heating. Some 5% of tiles have been subjected to abnormal loads at points distant from the contact area. Damage shows strong correlation of localised heating of the limiter with toroidal field ripple. Edge heating rates of 260 Mw/m 2 have not caused gross mechanical failure of the limiter. The mechanical damage comprises fatigue cracks analogous to those due to sustained loading at low power levels.

Deposition of carbon and beryllium and retention of deuterium on probes in the scrape-off layer of JET

Surface probes have been exposed at JET with all-carbon walls, when beryllium was evaporated onto graphite limiters, and with beryllium limiters. The aims have been to study the impurity content in the scrape-off, erosion and deposition, and the pumping of deuterium by codeposition with carbon and beryllium. Thermally activated erosion makes it difficult to study time resolved erosion close to the LCFS. In the all-carbon machine, the carbon density in the outboard scrapeoff between the belts was estimated to ~ 20% of the deuterium. With a fresh Be coating on the limiters, carbon in the SOL was initially suppressed, but increased from discharge to discharge. The D/C ratio in codeposition layers was three times larger with Be-evaporation than in all-carbon conditions. Edge profiles determined from impurity deposition became flatter during RF heating and current ramp down.

Global influx and impurity behaviour during ICRF heating in JET with beryllium gettering and beryllium limiters

During ICRF heating in JET with Be gettering and Be-belt limiters the impurity influxes are significantly reduced as compared to the all carbon case. The dominating “global” influxes originate at the limiters and wall and are consequence of the increased power flow in the scrape-off layer (SOL). The “RF specific” influxes from the torus internal surfaces, far away from the antenna, are not observed. Z eff = 2-1.5 is routinely achieved at power levels up to P rf = 15 MW and with Be limiters even at densities 〈 n e〉 = (1−2)× 10 19 m −3. The dilution is n d/ n e > 0.8. Z eff scales inversely with density and weakly with RF power, ( Z eff — 1) ∝ P 0.4–0.6 RF/〈 n e〉 0.9–1.1. P rad/ P tot is 50% lower then previously observed. Even at densities 〈 n e〉 = (4–6) × 10 19 m −3 the radiated fraction is below 30% of input power. In the case of Be gettering and carbon limiters the dominant impurity is carbon. With the Be limiters it is replaced by Be. The oxygen is eliminated in both cases. The density is controlled by the pumping of deuterium and the recycling coefficient R d is less than one during the entire rf pulse. The density limit observed in the all carbon phase was exceeded by a factor 2–4. Volume averaged density 〈 n e〉 = 1 × 10 20 m −3 was achieved with gas puffing at power P RF = 8 MW.

Global modelling of impurities in JET belt limiter discharges

The rate of production of impurities by physical sputtering is discussed in terms of a self-consistent zero-dimensional model based on energy balance. The importance of the effective self-sputter yield is emphasised and the effect of the uncertainty in this parameter is examined. The model is used to calculate how the plasma edge temperature, impurity influx, impurity concentration and total radiation depend on plasma average density and total input power. These results are compared with experimental results obtained using graphite and beryllium limiters in JET.

Results of JET operation with beryllium

The 1989 JET experimental campaign assessed the properties of beryllium as a plasma-facing material. This campaign had three phases: (i) an all graphite phase to recommission the tokamak, (ii) a phase with beryllium evaporation on graphite to isolate the effect of oxygen gettering and (iii) a beryllium limiter phase. Beryllium evaporation removed oxygen from the plasma and reduced carbon concentrations by a factor of two. Strong deuteron pumping was obtained which resulted in a significant, long-term improvement in density control and eliminated the deconditioning effect of disruptions. The use of beryllium limiters reduced the plasma radiation, increased the density limit, reduced the incidence of disruptions and enhanced the deuteron pumping. In the beryllium phases impurity influxes were lower during ICRH heating which permitted the first ICRH-only H-modes. The deuteron pumping allowed heavy gas fuelling to be used to reduce impurity levels of limiter and X-point plasmas.

Scrape-off layer parameters in JET with beryllium limiters

Scaling of edge particle flux and temperature with central density in JET has been studied in ohmic and auxiliary heated beryllium limiter discharges, using Langmuir probes in the edge. Data has been obtained over a much wider density range than previously, and it is found that flux density scales almost linearly with central density. The edge temperature, on the other hand, is almost constant at low densities, but falls rapidly when gas puffing is used in the edge to achieve high densities. A global power balance assuming T i = T e suggests a shortfall in power to the edge, particularly at low densities, suggesting T i > T e. The consequence of high T i on limiter sputtering is considered.

Effect of beryllium on plasma performance in JET

JET is investigating beryllium as material for walls and limiters. Studies were carried out with initially a thin layer of beryllium evaporated onto the walls of the machine, later in addition the graphite material of the belt limiter was exchanged against beryllium. The use of this material was generally beneficial for the plasma behaviour. Combined with a reduction in the oxygen content, strong pumping of hydrogen isotopes was found which allowed JET to widen considerably the operational space with respect to ion temperatures, densities and plasma purity. In this paper a comparison of the plasma performance with graphite and beryllium will be presented. The author discusses especially the impurity behaviour with respect to fluxes, concentrations, effective charge and dilution, and reports on density limits, disruption behaviour, wall pumping, hydrogen isotope retention, and the power handling capability of the beryllium limiter in the present design. Examples of improved plasma performance are given.

Beryllium release and beryllium transport in the tokamak UNITOR

The release and transport of beryllium in a tokamak plasma are investigated numerically in order to characterize the impurity concentrations and fluxes. A model has been developed and applied to the small-scale tokamak UNITOR when operated with beryllium limiters. The source term for the beryllium influx is evaluated to be approximately 10 18 s −1. The spatial distributions of the various beryllium ions are given which reveal that fully stripped beryllium is dominant in the plasma core. There, the contribution of beryllium to Z eff is at most 1. Further, the solution of rate equations indicates that the population densities of ground state and metastable neutral beryllium are of comparable order. These theoretical results will be used to interpret measurements of the beryllium concentrations and fluxes in front of the limiter by laser-induced fluorescence, which are now underway.

Asymmetries in the density at the edge of a tokamak plasma

Several Langmuir probes were used simultaneously to observe density asymmetries at the edge of the beam-heated tokamak ISX-B. The large number of identical discharges required for the joint JET/ISX-B beryllium limiter experiment offered the opportunity to perform to perform a careful edge study of the scrape-off layer using probes at various poloidal, toroidal, and radial locations. Floating, double, Langmuir probes were used at 3 separate locations. Edge density asymmetries were observed and correlated with the local connection length along a field line between material surfaces in the scrape-off layer. The effects of limiter movement were also explored with limiters other than the main limiter. It was observed that the asymmetries were altered somewhat when either the magnetic field or the plasma current was reversed. These spatial variations were accentuated locally when a certain critical density was exceeded and may be due to enhanced recycling at the main limiter just below the disruption-limited maximum density.

Thermal fatigue tests of a prototype beryllium limiter for jet

Beryllium is an attractive alternative to graphite for use as armor material for plasma interactive components in fusion devices because of its low atomic number, high strength, and good compatibility with hydrogen. However, beryllium is susceptible to damage from cyclic thermal stresses because of its high elastic modulus and thermal expansion coefficient. We have performed 2-D elastic-plastic finite element stress analyses of prototype beryllium limiter tiles for the JET project that are exposed to a surface heat flux of 250 W/cm2 for 15 second pulses. Plastic deformation was predicted to occur at the heated surface during both the heating and cooling phases of the cycle, thereby causing cyclic plastic strain. We also performed thermal fatigue tests using a rastered electron beam to apply the heat load to prototype limiter specimens. After 10 000 thermal fatigue cycles, the only damage of the beryllium tile was microcracking of the heated surface. The depth of this microcracking, 4 mm, corresponds closely to the calculated depth of cyclic plastic strain. These favorable results show that the operating conditions for the JET limiter design can be extended into the regime of cyclic plastic deformation without causing overall structural failure, despite the formation of thermal fatigue cracks.

Cleanup and gettering during the Beryllium Limiter Experiment in Impurity Study Experiment-B

Changes in plasma impurities and radiated power due to cleanup and gettering during the Beryllium Limiter Experiment are related to changes of the surface composition of the torus walls and residual gases in the torus. Pulse discharge cleaning was very effective in cleaning the ISX-B vacuum vessel after installation of the beryllium limiter. As a result of discharge cleaning, areas near the limiter were soon covered with a beryllium rich deposit containing some iron, nickel, chromium, and titanium as well as oxygen, carbon, sulfur, and nitrogen. Oxygen levels on the wall were reduced to below 10 at. %. Materials transport during discharge cleaning was apparently due to sputtered neutrals. Evaporation of beryllium from the limiter and subsequent deposition on the torus walls caused by high power tokamak shots produced gettering of the machine equal or better than that previously obtained with titanium or chromium. This gettering by beryllium eliminated the need for further discharge cleaning or other gettering. Beryllium transport in this case appears to be mostly as ions in the plasma edge.

Migration of beryllium in the Impurity Study Experiment-B tokamak during the Beryllium Limiter Experiment

A beryllium limiter was installed in the Impurity Study Experiment-B (ISX-B) tokamak to test the applicability of beryllium as a plasma-side component for possible use in the Joint European Torus (JET). This test was a collaborative effort involving both ORNL and JET personnel. The ISX-B tokamak was operated for more than 3000 high power neutral beam injected discharges using the beryllium limiter as the main confining aperture. In order to access the migration of beryllium in the tokamak and attached diagnostics, both stainless steel and silicon samples were placed at various strategic positions and used as catcher plates. Nuclear reaction analysis was used to measure the beryllium deposited on the plates and provided a measure of the beryllium migration. Both particulates and a thin film were found. The toroidal distribution of these two migration components is presented as well as a discussion of the analysis technique for the deposited beryllium. Analysis was made of the distribution of beryllium particulate matter deposited on the inside of the vessel wall. Both a size distribution and spatial distribution are presented.

Summary Abstract: Postmortem near surface analysis of beryllium limiter tiles from the Impurity Study Experiment-B

Summary Abstract: Postmortem near surface analysis of beryllium limiter tiles from the Impurity Study Experiment-B

Summary Abstract: Cleanup and gettering during the Beryllium Limiter Experiment in the Impurity Study Experiment-B

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Weight change measurements of erosion/deposition at beryllium limiter tiles in the Impurity Study Experiment-B

The weight changes of Be tiles which functioned as a rail limiter in ISX-B for more than 3500 beam-heated discharges have been determined. The net weight loss for the limiter was 2.0 g, with the central tiles losing a total of 3.2 g, and the inboard tiles gaining 1.2 g. The weight loss is attributed primarily to the release of Be droplets as a result of limiter surface melting. The weight gains resulted from an inward flow of molten material along the limiter surface. The results indicate high erosion (melt loss) with incomplete and nonuniform redeposition (melt flow) of limiter material during periods of limiter melting.

Thermal fatigue tests of a prototype beryllium limiter for jet

Beryllium is an attractive alternative to graphite for use as armor material for plasma interactive components in fusion devices because of its low atomic number, high strength, and good compatibility with hydrogen. However, beryllium is susceptible to damage from cyclic thermal stresses because of its high elastic modulus and thermal expansion coefficient. We have performed 2-D elastic-plastic finite element stress analyses of prototype beryllium limiter tiles for the JET project that are exposed to a surface heat flux of 250 W/cm 2 for 15 second pulses. Plastic deformation was predicted to occur at the heated surface during both the heating and cooling phases of the cycle, thereby causing cyclic plastic strain. We also performed thermal fatigue tests using a rastered electron beam to apply the heat load to prototype limiter specimens. After 10 000 thermal fatigue cycles, the only damage of the beryllium tile was microcracking of the heated surface. The depth of this microcracking, 4 mm, corresponds closely to the calculated depth of cyclic plastic strain. These favorable results show that the operating conditions for the JET limiter design can be extended into the regime of cyclic plastic deformation without causing overall structural failure, despite the formation of thermal fatigue cracks.

Impurity fluxes and concentrations in the ISX-B beryllium limiter experiment

An experiment to test beryllium as a limiter material has been performed in the ISX-B tokamak. One facet of the studies centred on characterizing impurity influxes and concentrations. The radiation from both low-Z (Be, C, N, O) and intermediate-Z (Ti, Cr) contaminants was measured at the limiter surface and at a wall location 90° away from the limiter to assess the relative contributions of metallic impurities from the two different sources. The effect of limiter melting, with concomitant reduction of both low-Z and intermediate-Z elements, was also documented. The analysis, including atomic rate coefficients, for interpreting spectral radiation in terms of production rates is discussed in detail.

Thermomechanical Testing of Beryllium for Limiters in ISX-B and JET

Materials testing of S-65-B grade beryllium has been conducted in order to evaluate the use of this material for limiter surfaces in the ISX-B and JET tokamaks. Selected thermal and mechanical properties were measured at temperatures up to 700 °C. These measurements revealed that S-65-B has exceptionally high ductility (up to roughly 50% elongation) at temperatures expected in normal operation of a beryllium limiter. Thermal fatigue tests under conditions relevant to limiters in ISX-B and JET were also performed using the Sandia National Laboratories Electron Beam Test System (EBTS). Results from these tests were compared to calculated results based on elastic-plastic finite element stress analyses. It was concluded from these tests and analyses that properly designed beryllium limiters should survive normal operation in ISX-B, JET, and similar devices without serious structural failure. Some degree of surface cracking can be expected, however, unless cyclic plastic deformation at the heated surfaces can be adequately controlled by careful design of the limiter.

Technical aspects of the joint JET-ISX-B beryllium limiter experiment

An experiment has been performed on the Impurity Study Experiment (ISX-B) tokamak to test beryllium as a limiter material. Beryllium is an attractive candidate for a limiter and has been proposed for use in the Joint European Torus (JET) experiment. A temperature-controlled, segmented, beryllium top-rail limiter was located inside the plasma radius described by the existing titanium limiters. An extended set of diagnostics was added for measurement of scrapeoff and limiter parameters. These included visible and infrared monitoring systems, probes, and surface analysis experiments. Tokamak experiments included parameter surveys of both Ohmically heated and neutral-beam-heated plasmas and an extended fluence test of the limiter. The most significant effect of operation with a beryllium limiter was the reduction in low-Z impurities caused by gettering action of beryllium deposited on the liner walls. The experiment required the design and implementation of contamination control apparatus and work procedures to prevent the accidental dispersion of beryllium dust.