Comprehensive nuclear materials 4 19 beryllium as a plasma facing material for near term fusion devices

Comprehensive nuclear materials 4 19 beryllium as a plasma facing material for near term fusion devices Comprehensive nuclear materials 4 19 beryllium as a plasma facing material for near term fusion devices Comprehensive nuclear materials 4 19 beryllium as a plasma facing material for near term fusion devices Comprehensive nuclear materials 4 19 beryllium as a plasma facing material for near term fusion devices Comprehensive nuclear materials 4 19 beryllium as a plasma facing material for near term fusion devices Comprehensive nuclear materials 4 19 beryllium as a plasma facing material for near term fusion devices

4.19 Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices G Federici Fusion for Energy, Garching, Germany R Doerner University of California at San Diego, San Diego, CA, USA P Lorenzetto Fusion for Energy, Barcelona, Spain V Barabash ITER Organization, St Paul Lez Durance, France ß 2012 Fusion for Energy (F4E) Published by Elsevier Ltd All rights reserved 4.19.1 4.19.2 4.19.2.1 4.19.2.2 4.19.2.3 4.19.3 4.19.3.1 4.19.3.1.1 4.19.3.1.2 4.19.3.1.3 4.19.3.1.4 4.19.3.2 4.19.3.2.1 4.19.3.2.2 4.19.3.3 4.19.3.3.1 4.19.3.3.2 4.19.4 4.19.4.1 4.19.4.1.1 4.19.4.1.2 4.19.4.2 4.19.4.3 4.19.4.4 4.19.4.4.1 4.19.4.4.2 4.19.4.4.3 4.19.4.4.4 4.19.4.4.5 4.19.5 4.19.5.1 4.19.5.1.1 4.19.5.1.2 4.19.5.2 4.19.6 4.19.6.1 4.19.6.1.1 4.19.6.1.2 4.19.6.2 Introduction Background Synopsis of PWIs in Tokamaks Brief History of Plasma-Facing Materials in Fusion Devices Experience with Beryllium in Tokamaks Beryllium PWI Relevant Properties Beryllium Erosion Properties Physical sputtering of beryllium Mixed-material erosion Chemically assisted sputtering of beryllium Enhanced erosion at elevated temperatures Hydrogen Retention and Release Characteristics Implantation Beryllium codeposition Mixed-Material Effects Be–C phenomena Be–W alloying Main Physical and Mechanical Properties General Considerations Physical properties Mechanical properties Selection of Beryllium Grades for ITER Applications Considerations on Plasma-Sprayed Beryllium Neutron-Irradiation Effects Thermal conductivity Swelling Mechanical properties Thermal shock effects Bulk tritium retention Fabrication Issues Joining Technologies and High Heat Flux Durability of the Be/Cu Joints Be/Cu alloy joining technology High heat flux durability of unirradiated Be/Cu joints Thermal Tests on Neutron-Irradiated Joints Tokamak PFC Design Issues and Predictions of Effects in ITER During Operation PFC Design Considerations Design of the beryllium ITER-like wall at JET Design of the beryllium ITER wall Predictions of Effects on the ITER Beryllium Wall During Operation 623 624 624 626 628 629 629 629 631 632 632 633 633 635 637 637 637 638 638 639 639 640 642 643 643 643 644 644 644 644 644 645 646 648 650 650 650 651 653 621 622 Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices 4.19.6.2.1 4.19.6.2.2 4.19.6.3 4.19.7 References Safety issues in ITER Erosion/damage of the ITER Be wall Prospect of Using Beryllium in Beyond-ITER Fusion Reactors Concluding Remarks Abbreviations Alcator C-Mod ANFIBE ASDEXUpgrade ATC CFC CIP CP DIII-D DIMES DS-Cu EAST The name Alcator was given to a class of tokamaks designed and built at the Massachusetts Institute of Technology; these machines are distinguished by high magnetic fields with relatively small diameters The high magnetic field helps create plasmas with relatively high current and particle densities The present incarnation is Alcator C-Mod Computer code for ANalysis of Fusion Irradiated BEryllium Axially Symmetric Divertor Experiment The original ASDEX, located in Garching, Germany, and decommissioned in about 1990, would qualify today as a medium sized tokamak It was designed for the study of impurities and their control by a magnetic divertor Its successor, ASDEX-Upgrade (a completely new machine, not really an ‘upgrade’), is larger and more flexible Adiabatic Toroidal Compressor Carbon-fiber composite Cold isostatic pressing Cold pressing A medium-sized tokamak, but the largest tokamak still operational in the United States Operated by General Atomics in San Diego Divertor Material Evaluation Studies, a retractable probe that allows the insertion and retraction of test material samples to the DIIID divertor floor, for example, for erosion/deposition studies Dispersion-strengthened copper Experimental advanced superconducting tokamak – an experimental superconducting ELMs FISPACT FZJ HIP INEEL ISX ITER JET 653 654 659 659 662 tokamak magnetic fusion energy reactor in Hefei, the capital city of Anhui Province, in eastern China Edge localized modes Inventory code included in the European Activation System Forschungszentrum Juelich, Germany Hot isostatic pressing Idaho National Engineering and Environmental Laboratory Now Idaho National Laboratory (INL) Impurity study experiment (ISX-A and ISX-B where two tokamaks operated at Oak Ridge National Laboratory) ITER, the world’s largest tokamak experimental facility being constructed in the South of France to demonstrate the scientific and technical feasibility of fusion power The project is being built on the basis of an international collaboration between the European Union, China, India, Japan, Russia, South Korea, and the United States The international treaty was signed in November 2006 and the central organization established in Cadarache Most of the components will be provided in kind by agencies set up for this purpose in the seven partners Joint European Torus – a large tokamak located at the Culham Laboratory in Oxfordshire, England, jointly owned by the European Community First device to achieve >1 W of fusion power, in 1991, and the machine that has most closely approached Q ¼ for DT operation (Q ¼ 0:95 in 1997) Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices JUDITH Juelich Divertor Test Facility in Hot Cells KSTAR Korea Superconducting Tokamak Advanced Reactor – a long-pulse, superconducting tokamak built in South Korea to explore advanced tokamak regimes under steady state conditions LANL Los Alamos National Laboratory LCFS Last closed flux surface MAR ITER Materials Assessment Report MIT Massachusetts Institute of Technology MPH ITER Materials Properties Handbook NBI Neutral beam injection NRA Nuclear reaction analysis NRI Nuclear Research Institute in the Czech Republic PDX Poloidal divertor experiment PFCs Plasma-facing components PISCES Plasma Interaction with Surface Components Experimental Station It is a plasma simulator located at the University of California San Diego in the United States (originally at University of California, Los Angeles) that is used to test materials and measure sputtering, retention, etc expected in tokamaks PLT Princeton Large Torus PWIs Plasma–wall interactions RES Radiation enhanced sublimation RMP Resonance magnetic perturbation SNL Sandia National Laboratory SOL Scrape-off-layer ST Symmetric tokamak (in this chapter) STEMET 1108 Brazing alloy: Cu–Sn–In–Ni TFR Torus Fontenay-aux-Roses TPE Tritium plasma experiment TRIM Transport of ion in matter code UCSD University of California, San Diego UNITOR One of the first small tokamaks where beryllium was used UTIAS University of Toronto Institute for Aerospace Studies VDE Vertical displacement event VHP Vacuum hot pressing 623 4.19.1 Introduction Beryllium, once called ‘the wonder metal of the future,’ is a low-density metal that gained early prominence as a neutron reflector in weapons and fission research reactors It then found a wide range of applications in the automotive, aerospace, defense, medical, and electronic industries Also, because of its unique physical properties, and especially favorable plasma compatibility, it was considered and used in the past for protection of internal components in various magnetic fusion devices (e.g., UNITOR, ISX-B, JET) Most important future (near-term) applications in this field include (1) the installation of a completely new beryllium wall in the JET tokamak, which has been completed by mid of 2011 and consists of $1700 solid Be tiles machined from t of beryllium; and (2) ITER, the world’s largest experimental facility to demonstrate the scientific and technical feasibility of fusion power, which is being built in Cadarache in the South of France ITER will use beryllium to clad the first wall ($700 m2 for a total weight of about 12 t of Be) Although beryllium has been considered for other applications in fusion (e.g., as neutron multiplier in the design of some types of thermonuclear breeding blankets of future fusion reactors and for hohlraums in inertial confinement fusion), this chapter will be limited to discussing the use of beryllium as a plasma-facing material in magnetic confinement devices, and in particular in the design, research, and development work currently underway for the JET and the ITER tokamaks Considerations related to health and safety procedures for the use of beryllium relevant for construction and operation in tokamaks are not discussed here Designing the interface between a thermonuclear plasma and the surrounding solid material environment has been arguably one of the greatest technical challenges of ITER and will continue to be a challenge for the development of future fusion power reactors The interaction between the edge plasma and the surrounding surfaces profoundly influences conditions in the core plasma and can damage the surrounding material structures and lead to long machine downtimes for repair Robust solutions for issues of plasma power handling and plasma–wall interactions (PWIs) are required for the realization of a commercially attractive fusion reactor A mix of several plasma-facing materials is currently proposed in ITER to optimize the requirements of areas with different power and particle flux characteristics 624 Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices (i.e., Be for the first wall, carbon-fiber composite (CFC) for the divertor strike point tiles, and W elsewhere in the divertor) Inevitably, this is expected to lead to cross-material contamination and the formation of material mixtures, whose behavior is still uncertain and requires further investigation The use of beryllium for plasma-facingcomponent (PFC) applications has been the subject of many reviews during the last two decades (see, e.g., Wilson et al.2 and Raffray et al.3 and references therein) Much of this fusion-related work has been summarized in a series of topical workshops on beryllium that were held in the past, bringing together leading researchers in the field of beryllium technology and disseminating information on recent progress in the field.4 Comprehensive reviews have also appeared recently in specialized journals5,6 containing state-of-the-art information on a number of topics such as manufacturing and development of coating techniques, component design, erosion/deposition, tritium retention, material mixing and compatibility problems, safety of beryllium handling, etc This chapter reviews the properties of beryllium that are of primary relevance for plasma protection applications in near-term magnetic fusion devices (i.e., PWIs, thermal and mechanical properties, fabricability and ease of joining, chemical reactivity, etc.) together with the available knowledge on performance and operation in existing fusion machines Special attention is given to beryllium’s erosion and deposition, the formation of mixed materials, and the hydrogen retention and release characteristics that play an important role in plasma performance, component lifetime, and operational safety The status of the available techniques presently considered for joining the beryllium armor to the heat sink material of Cu alloys for the fabrication of beryllium-clad actively cooled components for the ITER first wall is briefly discussed together with the results of the performance and durability heat flux tests conducted in the framework of the ITER first-wall qualification programme The effects of neutron irradiation on the degradation of the properties of beryllium itself and of the joints are also briefly analyzed This chapter is organized as follows Section 4.19.2 provides some background information for the reader and briefly reviews (1) the problem of PWIs in tokamaks; (2) the history of plasma-facing materials in fusion devices and the rationale for choosing beryllium as the material for the first wall of JET and ITER; and (3) the experience with the use of beryllium in tokamaks to date Section 4.19.3 describes in detail the beryllium PWI-relevant properties such as erosion/deposition, hydrogen retention and release, and chemical effects such as material mixing, all of which influence the selection of beryllium as armor material for PFCs Section 4.19.4 briefly reviews a limited number of selected physical and mechanical properties of relevance for the fabrication of heat exhaust components and the effects of neutron irradiation on material properties Section 4.19.5 describes the fabrication issues and the progress of joining technology and high heat flux durability of beryllium-clad PFCs Section 4.19.6 describes the main issues associated with the JET and ITER first-wall designs and discusses some constraints foreseen during operation The prospects of beryllium for applications in fusion reactors beyond ITER are briefly discussed Finally, a summary is provided in Section 4.19.7 4.19.2 Background 4.19.2.1 Synopsis of PWIs in Tokamaks A detailed discussion on this subject is beyond the scope of this review The relevant PWIs are comprehensively reviewed by Federici et al.7,8 More recent interpretations of the underlying phenomena and impact on the ITER device can be found in Roth et al.9 Here we summarize some of the main points PWIs critically affect tokamak operation in many ways Erosion by the plasma determines the lifetime of PFCs, and creates a source of impurities, which cool and dilute the plasma Deposition of material onto PFCs alters their surface composition and, depending on the material used, can lead to long-term accumulation of large in-vessel tritium inventories This latter phenomenon is especially exacerbated for carbonbased materials but there are still some concerns with beryllium Retention and recycling of hydrogen from PFCs affects fuelling efficiency, plasma density control, and the density of neutral hydrogen in the plasma boundary, which impacts particle and energy transport The primary driver for the interactions between the core plasma, edge plasma, and wall is the power generated in the plasma core that must be handled by the surrounding structures Fusion power is obtained by the reaction of two hydrogen isotopes, deuterium (D) and tritium (T), producing an a-particle and a fast neutron Although the kinetic energy carried by the 14.1 MeV neutron escapes the plasma and could be converted in future reactors beyond ITER to thermal energy in a surrounding blanket system, the Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices kinetic energy of the a-particle is deposited in the plasma The fraction of this power that is not radiated from the plasma core as bremsstrahlung or line radiation (and that on average is distributed uniformly on the surrounding structures) is transported across field lines to the edge plasma and intersects the material surfaces in specific areas leading to intense power loads The edge plasma has a strong influence on the core plasma transport processes and thereby on the energy confinement time A schematic representation of the regions of the plasma and boundary walls in a divertor tokamak is portrayed in Figure taken from Federici et al.7 The outermost closed magnetic field surface forms an X-point of zero poloidal magnetic field within the vessel This boundary is called the ‘last closed flux Magnetic flux surfaces Separatrix (LCFS) Edge region Scrape-off layer First wall Plasma core Separatrix (LCFS) X-point Divertor region Baffle Vertical divertor target plate Private flux region Separatrix strike point Pump Figure Poloidal cross-section of a tokamak plasma with a single magnetic null divertor configuration, illustrating the regions of the plasma and the boundary walls where important PWIs and atomic physics processes take place The characteristic regions are (1) the plasma core, (2) the edge region just inside the separatrix, (3) the scrape-off-layer (SOL) plasma outside the separatrix, and (4) the divertor plasma region, which is an extension of the SOL plasma along field lines into the divertor chamber The baffle structure is designed to prevent neutrals from leaving the divertor In the private flux region below the X-point, the magnetic field surfaces are isolated from the rest of the plasma Reproduced with permission from Federici, G.; Skinner, C H.; Brooks, J N.; et al Plasma-material interactions in current tokamaks and their implications for next-step fusion reactors Nucl Fusion 2001, 41, 1967–2137 (review special issue), with permission from IAEA 625 surface’ (LCFS) or ‘separatrix.’ Magnetic field surfaces inside the LCFS are closed, confining the plasma ions The edge region, just inside the LCFS, contains significant levels of impurities not fully ionized, and perhaps neutral particles Impurity line radiation and neutral particles transport some power from here to the wall The remaining power enters the region outside the LCFS either by conduction or, depending on the degree to which neutrals penetrate the plasma, by convection This region is known as the scrape-offlayer (SOL) as here power is rapidly ‘scraped off ’ by electron heat conduction along open field lines, which are diverted to intersect with target regions that are known as ‘divertors.’ Poloidal divertors have been very successful at localizing the interactions of plasma ions with the target plate material in a part of the machine geometrically distant from the main plasma where any impurities released are well screened from the main plasma and return to the target plate The plasma density and temperature determine the flux density and energy of plasma ions striking the plasma-wetted surfaces These, in turn, determine the rate of physical sputtering, chemical sputtering, ion implantation, and impurity generation The interaction of the edge plasma with the surrounding solid material surfaces is most intense in the vicinity of the ‘strike point’ where the separatrix intersects the divertor target plate (see inset in Figure 1) In addition, the plasma conditions determine where eroded material is redeposited, and to what degree codeposition of tritium occurs at the wall The plasma power flow also determines the level of active structural cooling required Typical plasma loads and the effects expected during normal operation and off-normal operation in ITER are summarized in Table Because of the very demanding power handling requirements (predicted peak value of the heat flux in the divertor near the strike-points is >10 MW mÀ2) and the predicted short lifetime due to sputtering erosion arising from very intense particle fluxes ($1023–1024 particles mÀ2 sÀ1) and damage during transient events, beryllium has been excluded from use in the ITER divertor and is instead the material selected for the main chamber wall of ITER Recent observations in present divertor tokamaks have shown that plasma fluxes to the main wall are dominated by intermittent events leading to fast plasma particle transport that reaches the PFCs along the magnetic field (see Loarte et al.10 and references therein) The quasistationary heat fluxes to the main wall are thought to be dominated by convective 626 Table Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices Major issues associated with operation of ITER PFCs PFCs Divertor – strike-point regions Plasma loads Candidate armor Effects  Radiation and particle heat  Large particle fluxes  Disruptions  ELM’s  Slow-high power CFCa Chemical erosion evaporation brittle destruction and tritium codeposition  Radiation heat  Disruptions  Radiation heat (MARFE’sb)  $100 eV ions and CX  Erosion lifetime and component replacement  High tritium inventory and safety transients Divertor – baffle region Dome Issue W W High sputtering evaporation/ melting High sputtering evaporation/ melting  Plasma contamination  Erosion lifetime neutrals First wall Start-up limiters  Moderate power transients  Plasma contact during VDEsc  Disruptions and runaway electrons  ELMs  High start-up heat loads  Plasma contact during VDEs  Disruptions Be Evaporation/melting  T retention in beryllium codeposited layers  Chemical reactivity especially with Be dust Be High sputtering evaporation/ melting  Erosion lifetime a W is also considered as an alternative Multifaceted asymmetric radiation from the edge (MARFE) Vertical displacement event (VDE) b c transport,11 but still remain to be clarified Although the steady-state parallel power fluxes associated with these particle fluxes will only be of the order of several MW mÀ2 in the ITER QDT ¼ 10 reference scenario, local overheating of exposed edges of main wall PFCs can occur because of limitations in the achievable alignment tolerances Similarly, transient events are expected to cause significant power fluxes to reach first-wall panels in ITER along the field lines Edge localized modes (ELMs) deposit large amounts of energy in a short time, and in some cases in a toroidally localized fashion, which can lead to strong excursions in PFC surface temperatures While the majority of ELM energy is deposited on divertor surfaces, a significant fraction is carried to surfaces outside the divertor There are obvious concerns that ELMs will lead to damage of the divertor and the first wall.12 An additional concern is that even without erosion, thermal shock can lead to degradation of material thermomechanical properties, for example, loss of ductility leading to an enhanced probability of mechanical failure or spalling (erosion) Research efforts to characterize the ELMs in the SOL are described elsewhere.13–15 There are still large uncertainties in predicting the thermal loads of ELMs on the ITER beryllium first wall and the range of parallel energy fluxes varies from 1.0 MJ mÀ2 (controlled ELMs) to 20 MJ mÀ2 (uncontrolled ELMs).16,17 Even for controlled ELMs, such energy fluxes are likely to cause melting of up to several tens of micrometers of beryllium at the exposed edges,18 which could cause undesirable impurity influxes at every ELM.10,11 4.19.2.2 Brief History of Plasma-Facing Materials in Fusion Devices PWIs have been recognized to be a key issue in the realization of practical fusion power since the beginning of magnetic fusion research By the time of the first tokamaks in the 1960s in the USSR and subsequently elsewhere, means of reducing the level of carbon and oxygen were being employed.19,20 These included the use of stainless steel vacuum vessels and all-metal seals, vessel baking, and discharge cleaning Ultimately, these improvements, along with improved plasma confinement, led to the first production of relatively hot and dense plasmas in the T3 tokamak ($1 keVand $3 Â 1019 mÀ3).21,22 These plasmas, while being cleaner and with low-Z elements fully stripped in the core, still had unacceptable levels of carbon, oxygen, and metallic impurities The metallic contamination inevitably consisted of wall and limiter materials Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices Early in magnetic fusion research, it was recognized that localizing intense PWIs at some type of ‘sacrificial’ structure was desirable, if only to ensure that more fragile vacuum walls were not penetrated This led to the birth of the ‘limiter,’ usually made to be very robust, from refractory material and positioned to ensure at least several centimeters gap between the plasma edge and more delicate structures like bellows, electrical breaks, vacuum walls, etc Typical materials used for limiters in these early days included stainless steel in Adiabatic Toroidal Compressor (ATC)23 and ISX-A24 and many others, molybdenum in Alcator A25 and Torus Fontenay-aux-Roses (TFR),26 tungsten in symmetric tokamak (ST)27 and Princeton Large Torus (PLT),28 and titanium in poloidal divertor experiment (PDX).29 Poloidal divertors have been very successful at localizing the interactions of plasma ions with the target plate material in a part of the machine geometrically distant from the main plasma where any impurities released are well screened from the main plasma and return to the target plate.30 By the early 1980s, it was also recognized that in addition to these functions, the divertor should make it easier to reduce the plasma temperature immediately adjacent to the ‘limiting’ surface, thus reducing the energies of incident ions and the physical sputtering rate Complementary to this, high divertor plasma and neutral densities were found The high plasma density has several beneficial effects in dispersing the incident power, while the high neutral density makes for efficient pumping Pumping helps with plasma density control, divertor retention of impurities and, ultimately, in a reactor, helium exhaust By the late 1970s, various tokamaks were starting to employ auxiliary heating systems, primarily neutral beam injection (NBI) Experiments with NBI on PLT resulted in the first thermonuclear class temperatures to be achieved.28,31,32 PLT at the time used tungsten limiters, and at high powers and relatively low plasma densities, very high edge plasma temperatures and power fluxes were achieved This resulted in tungsten sputtering and subsequent core radiation from partially stripped tungsten ions For this reason, PLT switched limiter material to nuclear grade graphite Graphite has the advantage that eroded carbon atoms are fully stripped in the plasma core, thus reducing core radiation In addition, the surface does not melt if overheated – it simply sublimes This move to carbon by PLT turned out to be very successful, alleviating the central radiation problem For these reasons, carbon has tended to be the favored limiter/divertor material in magnetic fusion research ever since 627 By the mid-1980s, many tokamaks were operating with graphite limiters and/or divertor plates In addition, extensive laboratory tests and simulations on graphite had begun, primarily aimed at understanding the chemical reactivity of graphite with hydrogenic plasmas, that is, chemical erosion Early laboratory results suggested that carbon would be eroded by hydrogenic ions with a chemical erosion yield of Y $ 0.1 C/Dỵ, a yield several times higher than the maximum physical sputtering yield Another process, radiation-enhanced sublimation (RES), was discovered at elevated temperatures, which further suggested high erosion rates for carbon Carbon’s ability to trap hydrogenic species in codeposited layers was recognized These problems, along with graphite’s poor mechanical properties in a neutron environment (which had previously been known for many years from fission research33), led to the consideration of beryllium as a plasma-facing material This was primarily promoted at JET.34 A description of the operation experience to date with Be in tokamak devices is provided in Section 4.19.2.3 At present, among divertor tokamaks, carbon is the dominant material only in DIII-D Alcator C-Mod at Massachusetts Institute of Technology (MIT), USA35 uses molybdenum ASDEX-Upgrade (Axially Symmetric Divertor Experiment) is fully clad with tungsten,36 and JET has completed in 2011 a large enhancement programme37 that includes the installation of a beryllium wall and a tungsten divertor New superconducting tokamaks, such as Korea Superconducting Tokamak Advanced Reactor (KSTAR) in Korea38 and experimental advanced superconducting tokamak (EAST) in China,39 employ carbon as material for the in-vessel components, but with provisions to exchange the material later on in operation The current selection of plasma-facing materials in ITER has been made by compromising among a series of physics and operational requirements, (1) minimum effect of impurity contamination on plasma performance and operation, (2) maximum operational flexibility at the start of operation, and (3) minimum fuel retention for operation in the DT phase This compromise is met by a choice of three plasma-facing materials at the beginning of operations (Be, C, and W) It is planned to reduce the choices to two (Be and W) before DT operations in order to avoid long-term tritium retention in carbon codeposits during the burning plasma phase Beryllium has been chosen for the first-wall PFCs to minimize fuel dilution caused by impurities released from these surfaces, which are expected to 628 Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices have the largest contamination efficiency.40–44 Moreover, the consequences of beryllium contamination on fusion performance and plasma operations are relatively mild This has been demonstrated by experiments in tokamaks (see Section 4.19.2.3) The main issues related to the use of beryllium in ITER are (1) the possible damage (melting) during transients such as ELMs, disruptions, and runaway electron impact, and its implications for operations and (2) the codeposition of tritium with beryllium which is eroded from the first wall and deposited at the divertor targets (and possibly also locally redeposited into shadowed areas of the shaped ITER first wall) Both issues are part of ongoing research, the initial results of which are being taken into account in the ITER design so that the influence of these two factors on ITER operation and mission is minimized This includes ELM control systems based on pellets and resonance magnetic perturbation (RMP) coils, disruption mitigation systems, and increased temperature baking of the divertor to release tritium from the beryllium codeposited layers Carbon is selected for the high power flux area of the divertor strike points because of its compatibility with operation over a large range of plasma conditions and the absence of melting under transient loads Both of these characteristics are considered to be essential during the initial phase of ITER exploitation in which plasma operational scenarios will require development and transient load control and mitigation systems will need to be demonstrated 4.19.2.3 Experience with Beryllium in Tokamaks Only three tokamaks have operated with beryllium as the limiter or first-wall material The first experiments were performed by UNITOR,45 which were then followed by ISX-B.46 Both tokamaks investigated the effects of small beryllium limiters on plasma behavior (UNITOR had side limiters at two toroidal locations and ISX-B had one top limiter) in support of the more ambitious beryllium experiment in JET (see below) The motivation to use beryllium came from the problem of controlling the plasma density and impurities when graphite was used Both UNITOR and ISX-B showed that once beryllium is evaporated from the limiter and coated over a large segment of the first wall, oxygen gettering leads to significant reduction of impurities When the heat load on the beryllium limiter was increased to the point of evaporating beryllium, the oxygen concentration was decreased dramatically Although the concentration of beryllium in the plasma was increased, its contribution to Zeff (the ion effective charge of the plasma Zeff provides a measure for impurity concentration) was more than compensated by the reduction of oxygen, carbon, and metal impurities.45 The plasma Zeff was observed to be reduced from 2.4 to near unity with beryllium It must be noted that there was a negative aspect associated with beryllium operation during the ISX-B campaign The reduction in plasma impurities was not observed until the limiter surface was partially melted causing beryllium to be evaporated and coated on the first wall Once melting did occur, the droplets made subsequent evaporation more likely but hard to control The consequent strong reduction in plasma impurities associated with gettering then made discharge reproducibility hard to obtain However, if a much larger plasma contact area is already covered with Be, one does not need to rely on limiter melting to obtain the beneficial effect of beryllium This effect could be achieved by using large area beryllium limiter, or coating the inside wall with beryllium which was the approach taken by JET when it introduced beryllium in 1989 Large tokamak devices such as JET had found it very difficult to control the plasma density with graphite walls as the discharge pulse length got longer Motivated by the frequent occurrence of a phenomenon that plagued the earlier campaigns – the so-called carbon blooms due to the overheating of poorly designed divertor tiles and the subsequent influx of carbon impurities in the plasma due to evaporation – JET decided to use beryllium as a plasma-facing material Thin evaporated beryllium layers on the graphite walls were used initially ($100 A˚ average thickness per deposition) on the plasma-facing surface of the device Subsequently, beryllium tiles were installed on the toroidal belt limiter The main experimental results with beryllium can be summarized as follows: The concentration of carbon and oxygen in the plasma were 4–7% and 0.5–2%, respectively, when graphite was used as belt limiter With a beryllium belt limiter, the carbon content was reduced to 0.5% and oxygen became negligible, because of oxygen gettering by beryllium During ohmically heated discharges, the concentration of beryllium remained negligible even though beryllium was the dominant impurity Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices While the value of Zeff was $3 using the graphite limiter and auxiliary heating power of 10 MW, Zeff was $1.5 even with additional heating powers of up to 30 MW with a beryllium limiter The fuel density control was greatly improved with the beryllium limiter and beryllium evaporated wall Gas puffing to maintain a given plasma density was typically 10 times larger when using beryllium than graphite Following the beryllium limiter experience, divertor beryllium targets were installed in JET for two configurations An extensive set of experiments with toroidally continuous X-point divertor plates was carried out in JET in the period 1990–1996 to characterize beryllium from the point of view of its thermomechanical performance, as well as its compatibility with various plasma operation regimes.47–50 In the JET Mk I experiments, melting of the beryllium tiles was reached by increasing (in a progressive way) the power flux to a restricted area of the divertor target in fuelled, medium density ELMy H-mode discharges (Pinp $ 12 MW) Large beryllium influxes were observed when the divertor target temperature reached $1300  C In these conditions, it became difficult to run low-density ELMy H-mode discharges (Pinp $ 12 MW) without fast strike point movement (to achieve lower average power load) and the discharges either had very poor performance or were disrupted However, no substantial plasma performance degradation was observed for medium density H-modes with fixed strike point position, or if fast strike point movement was applied in lowdensity H-modes, despite the large scale distortion of the target surface caused by the melt layer displacement and splashing due to the previous $25 high power discharges48,51 (see Figure 252) This demonstrated that it was possible to use the damaged Be divertor target as the main power handling PFC if the Figure Melting of the Joint European Torus Mk I beryllium target plate tiles after plasma operation Image courtesy of EFDA-JET 629 average power load was decreased, either by increasing plasma density and radiative losses, or by strike point sweeping The damage did not prohibit subsequent plasma operation in JET, but would seriously limit the lifetime of Be PFCs in long-pulse ITER-like devices The latest results of the operation of JET with beryllium have been reviewed recently by Loarte et al.10 4.19.3 Beryllium PWI Relevant Properties This section describes the present understanding of PWIs for beryllium-containing surfaces First, it focuses on the erosion properties of ‘clean’ beryllium surfaces at different temperatures Retention of plasma fuel species in both bulk and codeposited layers of beryllium is then described As beryllium will not be used as the exclusive plasma-facing material in future confinement devices, issues associated with mixed, beryllium-containing surfaces are also addressed 4.19.3.1 Beryllium Erosion Properties The term erosion is used to describe a group of processes that remove material from a surface subjected to energetic particle bombardment Included under the general classification of erosion are processes such as physical sputtering, chemically assisted physical sputtering, chemical sputtering, and thermally activated release from surfaces Of these processes, only chemical sputtering, where volatile molecular species are formed on the surface, appears to be inactive in beryllium 4.19.3.1.1 Physical sputtering of beryllium Physical sputtering results from the elastic transfer of energy from incoming projectiles to atoms on the surface of the target material Target atoms can be sputtered when the energy they receive from the collisional cascade of the projectile exceeds the binding energy of the atom to the surface The physical sputtering rate is usually referred to as the sputtering yield, Y, which is defined as the ratio of the number of atoms lost from a surface to the number of incident energetic particles striking the surface The lower the binding energy of surface atoms, the larger the physical sputtering yield As physical sputtering can be approximated using a series of binary collisions within the surface, it is relatively easy to estimate Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices the physical sputtering yield of given projectile-target scenarios Monte-Carlo based simulation codes (such as transport of ions in matter (TRIM))53 have been used to generate extensive databases of sputtering yields based on incident particle angle, energy, and mass, for a variety of targets54 including beryllium Measurement of the physical sputtering yield from a beryllium surface is complicated by the natural affinity of beryllium for oxygen A beryllium surface will quickly form a thin, stable, passivating oxide surface layer when exposed to atmosphere In ion beam devices, it is possible to clean any oxides from the beryllium surface before a measurement and with careful control of the residual gas pressure, make the measurements before the oxide reforms on the surface and alters the measurement.55 It has also been shown that it is possible to deplete the beryllium surface of oxide by heating the sample to temperatures exceeding 500  C, where the beryllium below the oxide can diffuse through the oxide to the surface,56 thereby allowing measurements on a clean beryllium surface The comparison between the calculated sputtering yield and measurements made using mass-selected, monoenergetic ion-beams devices impinging on clean beryllium surfaces is fairly good.57 Measurements of sputtering yields in plasma devices, however, are complicated by several factors In plasma devices, the incident ions usually have a temperature distribution and may contain different charge state ions Each different charge state ion will be accelerated to a different energy by the electrostatic sheath in the vicinity of the surface When hydrogenic plasma interacts with a surface, one must also account for a distribution of molecular ions striking the surface In the case of a deuterium plasma, for example, the distribution of molecular ỵ ions (Dỵ, Dỵ , D3 ) must be taken into account as the incident molecule disassociates on impact with the surface and a Dỵ ion becomes equivalent to the bombardment of two deuterium particles with one-half the incident energy of the original Dỵ ion Figure shows the change to the calculated sputtering yield when one includes the effects of molecular ions in a plasma, compared to the calculated sputtering yield from pure Dỵ ion bombardment The trajectory of the incoming ions can also be altered by the presence of electrostatic and magnetic sheaths Plasmas also contain varying amounts of impurity ions, originating either from PWIs in other locations of the device, or ionization of residual background gas present in the device and these impurity ions, or simply neutral gas atoms, may interact with the surface Finally, the incident flux from the plasma is usually so large that the surface being investigated, and its morphology, becomes altered by the incident flux and a new surface exhibiting unique characteristics may result Nevertheless, the physical sputtering yield from beryllium surfaces exposed to plasma ion bombardment has been measured in several devices Unfortunately, there is little consensus on the correct value of the physical sputtering yield In JET, the largest confinement device to ever employ beryllium as a PFC sputtering yield measurements range from values far exceeding47 to values less58 than one would expect from the predictions of TRIM In the Plasma Interaction with Surface Components Experimental Station B (PISCES-B) device, systematic experiments to measure the physical sputtering yield routinely show values less59–61 than those expected from TRIM This difference is shown in Figure 3, where the energy dependence of the calculated yield is compared to experimental measurements Another primary difference between the conditions in an ion beam device and those encountered in a plasma device has to with the neutral density near the surface being investigated In an ion beam experiment, the background pressure is kept very low 0.1 Physical sputtering yield 630 0.01 0.001 0.0001 10-5 20 40 60 80 100 120 140 160 Ion energy (eV) Measured yield in D plasma Calculated yield (D+ ions only) Calculated yield (D+, D+2 , D+3 ions) Figure Calculated sputtering yields from pure Dỵ bombardment at normal incidence compared to that ỵ calculated for a (0.25, 0.47, 0.28) mix of Dỵ, Dỵ , and D3 ; also shown is the measured yield from such a plasma 652 Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices ‘Toast rack’ carrier Be slices Support pins Halo current path through the tile JG06.336-7 c Halo current from plasma Figure 18 Inner wall guard limiter tile (exploded view, top, and prototype, bottom) The five castellated Be slices have interslice and outer slice internal toroidal edges ski-slope shadowed The slices are held on an inconel carrier by pins which allow bowing under thermal load The RH bolts are designed to be shadowed by the next installed tile Reproduced with permission from Riccardo, V J Nucl Mater 2009, 390–391, 895–899 normal or near-normal field lines, emanating in the near SOL; (3) provide power load capability of 4–5 MW mÀ2, in order to be able to use the first wall as a limiter for startup and termination; and (4) withstanding transients is still subject to discussion In particular, it must be noted that the vulnerability to damage induced by thermal transients is recognized and linked to the feasibility and efficiency of all processes required for full remote maintenance of the first-wall panels, which is yet to be demonstrated In practical terms, the approach adopted is to provide a shadowed poloidal band in the center of the first-wall panel, the two sides being shaped in a form typical for limiters both to provide the shadowing of the band and to ensure that the toroidally facing edge of the first-wall panel is shadowed Because of the port regions on the low-field side, which contain a variety of structures with varying power handling capabilities, and because the toroidal field ripple is variable with the toroidal field, it is not possible to exploit the entire wall surface in this location For this reason, the firstwall panels on the low-field side have the poloidal bands between the ports advanced with respect to those in line with the ports (see Figure 19) The amount of set back required at the edges of the first wall is determined by the penetration angle of the field lines and the power scrape-off length, with the optimization taking into account the differing power handling capability of the front face and the edge of the first wall Considerations discussed here are limited to some problems associated with the design of the beryllium tiles and prediction of PWI effects during operation in ITER An important design driver for the first wall in the past was the specification of the thermal load during off-normal transient events.3 In particular, the thickness of the beryllium tiles had to be such as to prevent overheating of the joints and possible damage of the coolant pipes (see Section 4.19.6.2.2) Also, the thickness of the tile determines the temperature Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices Be wall CFC strikepoints W elsewhere (a) (b) Figure 19 (a) View of the low field side first-wall surface showing how the panels in line with the port openings are recessed with respect to those between It also shows the shielded central section of the panels allowing for access to the mechanical and hydraulic connections Reproduced from Hawryluk, R J.; et al Nucl Fusion 2009, 49, 15, 065012; with permission from IAEA (b) Allocation of armor materials Reproduced from Hawryluk, R J.; et al Nucl Fusion 2009, 49, 15, 065012; Federici, G.; Loarte, A.; Strohmayer, G Plasma Phys Contr Fusion 2003, 45, 1523–1547, with permission from IOPP gradient and the thermal stress under a prescribed thermal load during steady-state Limits on the tile temperature during operation arise as a result of many processes including melting, excessive vaporization, thermal fatigue, reduced mechanical integrity, and chemical reactions during accidental exposure of armor or structure to air or steam The last one of the above processes is important as explosion of hydrogen liberated from the steam–Be reaction is a major concern In the past, a tile thickness of 10 mm was adopted This corresponded to a Be maximum temperature limit of $650–750  C, roughly the level at which the relevant Be material properties (including mechanical, embrittlement, thermal fatigue, and swelling effects) start to degrade considerably Because of the differences in the product of the elastic modulus and the coefficient of thermal expansion (E) between beryllium and copper or copper alloys (EBe/ECuCrZr ¼ 2.4), large thermal stresses are set up around the bond between the beryllium tile and copper allot heat-sink The difficulty to successfully join low thermal expansion armor materials such as beryllium and tungsten to high thermal expansion heat sink materials has been a major problem and has been discussed in 653 Section 4.19.5 Thermomechanical modeling has shown the desirability of using very small tiles of brush like structure for PFC armor because of the reduction of the stress at the armor–heat sink interface The proper selection of the size of beryllium tile is an important issue which impacts all aspects of component manufacturing such as increased cost of machining, nondestructive examination features, reliability and repair of unbonded tiles, etc In general, the fatigue life issue is difficult to quantify because of a number of factors The thermal stresses depend on the temperature profile and the degree of constraint in the tiles Tile castellations must be introduced to further relieve the constraints, and these have been sized following an extensive program of coupled thermal and mechanical analyses using finite elements codes such as ANSYS185 and ABAQUS.186 4.19.6.2 Predictions of Effects on the ITER Beryllium Wall During Operation 4.19.6.2.1 Safety issues in ITER 4.19.6.2.1.1 In-vessel tritium inventory Estimates of the tritium inventory and of permeation in the PFCs of a magnetic fusion device are important for assessing the radiological hazards from routine operation and from potential accidents, for the design of the water detritiation system, and for predicting the tritium supply requirements In addition, these estimates have contributed to the decisions involving the choice of different armor materials in ITER options, which have a strong impact on tritium retention In spite of the experimental and modeling progress which has taken place in the recent past, understanding of the subject of tritium–wall interactions is still far from complete and quantification of the tritium inventory in ITER is highly uncertain The retention and permeation of implanted tritium in ITER PFCs have been widely studied in the past (see Section 4.19.3 and the example of calculations found elsewhere).9,187–189 On the basis of the results of these calculations, it can be concluded that the inventory of tritium in the beryllium first wall of a device like ITER, due to implantation, diffusion, trapping, and neutron-induced transmutation, will be on the order of 100 g rather than the kilogram quantities estimated previously70,100 and most of that will come from neutron-induced transmutations in the Be itself The dominant process for long-term retention of tritium in beryllium for ITER is expected to be 654 Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices codeposition (see Section 4.19.3.2.2) with eroded wall material (i.e., the incorporation of tritium in the deposited layers where impurity atoms or molecules are deposited together with eroded material and a flux of energetic or thermal atoms) The inventory of this potentially volatile tritium must be kept as low as reasonably achievable ($1 kg tritium), in order to minimize the impact on the environment in case of an accidental release, in particular to avoid the evacuation of the neighboring population The rate of formation of the codeposited material depends on the energy of the incident particles and on the substrate temperature during the deposition In ITER, the total amount of tritium trapped in the codeposited layers will strongly depend on whether carbon is retained in the divertor during DT operation But even in a full metal ITER configuration (e.g., with Be wall and W divertor) there is evidence for potential tritium accumulation for ITER in deposited Be layers.9 In contrast to carbon, tritium codeposition in beryllium layers is expected to be released at relatively low temperature and there are provisions to periodically bake the divertor in ITER at 350  C to release tritium trapped in the codeposited Be layers (see Section 4.19.3.2.2) 4.19.6.2.1.2 Chemical reactivity of beryllium dust with steam in ITER Although not a concern in present day tokamaks, in-vessel dust and tritium inventories have been recognized as a safety and operational issue for next step devices such as ITER.190–193 In particular, accident scenarios that result in water or steam exposure of hot plasma-facing materials are one of the greatest concerns for ITER, because steam interacts with hot beryllium leading to the production of hydrogen, and hydrogen in the presence of air can lead to an explosion The steam-chemical reactivity of different grades of Be has been studied extensively in the past.194–200 The amount of hydrogen produced depends on the specific material, temperature, exposure time, and especially the effective surface area Because of the large surface area of dust, its chemical reactivity is an issue Dust is expected to be produced inside the vacuum vessel of a tokamak by interaction of the plasma with the components of the first wall and the divertor A detailed discussion of the mechanisms of dust production and of the influence of parameter variations is beyond the scope of this contribution, but it should be noted that the processes and the production rate of dust are not fully understood and the extrapolation of knowledge from existing tokamaks to ITER is difficult Research into dust production mechanisms and rates, the appropriate dosimetric limits for personnel exposure, and methods of removal has only recently begun.201,202 The location where the dust settles will determine its temperature, and consequently, its chemical reactivity At the moment about kg of C, kg of W, and kg of Be dust are allowed ‘on hot surfaces’ in ITER, with these limits set by the H production risk This corresponds to the maximum allowable quantity of H (2.5 kg) for the vessel integrity to be guaranteed in case of explosion A complete oxidation of Be at 400  C and C at 600  C is assumed for the calculation If no C is present in the machine, the limits are relaxed to 11 kg for Be, or 230 kg for W These quantities are set such that the overall hydrogen combustion limit is not exceeded.9 It must be recognized that a limit on the order of $10 kg for beryllium dust on ‘hot-surfaces’ is very restrictive, and in particular, the development of diagnostics techniques that can determine from local measurements the global inventory in the machine could prove to be very challenging.203 However, it is also likely that dust in ITER produced by Be eroded from the wall and deposited on the divertor will not survive on plasma-facing surfaces exposed to heat fluxes and will tend to accumulate in grooves or castellations in the armors of PFCs They are an essential feature of the design of PFCs to relieve stresses during cyclic high heat flux loading, thus maximizing the fatigue lifetime of the armor to heat-sink joint Some reduction in reaction rates is expected because the steam supply is not unlimited and steam must diffuse through the dust in the grooves Experiments have been carried out in the Russian Federation, both in the Bochvar Institute of Moscow and the Efremov Institute of St Petersburg.204 Although not conclusive, the main results summarized in Figure 20, show a reduction of the measured Be steam reactivity, particularly at high temperatures (more than a factor of 20) However, further experimental and modeling work is needed to clarify if the observed slower kinetics at high temperatures (800–900  C) eliminates the risk of explosion in the event of an accident 4.19.6.2.2 Erosion/damage of the ITER Be wall The erosion mechanisms that affect the erosion/ damage of the first wall in ITER are (1) sputtering erosion by D–T ions and charge-exchange neutrals Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices 1200 1E + 02 1000 H2 generation rate (l m- s-1) 1E + 01 1E + 00 (3) 1E - 01 800 700 600 500 400 ºC (1) Pressed powder in grooves (Efremov Institute) (2) Nonpressed powder in grooves (Bochvar Institute) (3) Powder on open surfaces (Efremov Institute) (1) 1E - 02 1E - 03 (2) 1E - 04 655 1E - 05 INEL 91-GV-P INEEL 97-PSA-G INEL 92-GV-P INEEL 97-PSA-WG INEL 91-GV-PS INEL 96-GMS-C,D INEL 92-WG-DD INEL 96-GMS-I INEL 92-GV-D,D INEEL 97-PSB-G INEL 96-WG-D,D INEEL 97-PSB-WG INEL 96-GMS-D,D INEEL 97-RA-G2 INEL 96-WG-C,D INEEL 97-RA-WG2 INEL 96-WG-I INEEL 97-RA-G1 INEEL 97-RA-WG1 1E - 06 1E - 07 1E - 08 10 11 12 13 14 Inverse temperature ((1 K- 1) ´ 10 000) 15 16 17 Courtesy of V.Filatov (Efremov) Figure 20 Initial reactivity of Be powder in grooves with steam in recent experiments carried out at Efremov and Bochvar Institute in the Russian Federation (Be powder: BET-0.38 m2 gÀ1, average partial size ¼ 15 mm, free (nonpressed) dust density ¼ 0.7 g cmÀ3) during normal operation; and (2) evaporation and loss of melt layers during off-normal transient events such as thermal quench disruptions, ELMs, VDEs, and runaway electrons impact There are additional localized erosion phenomena such as arcing, overheating with evaporation, and, possibly, loss of melt layer on exposed edges, but it is very difficult to make predictions of these effects for ITER Special design attention has been given to avoid the misalignments of PFCs and avoid thermal overloading with possible localized damage 4.19.6.2.2.1 Erosion of Be wall during normal operation Calculations have been done to compute erosion of the first wall (due to fuel charge-exchange neutrals and ions, and impurity ions).205,206 It was found that about 20–40 g of Be per 400 s discharge are eroded from the wall with a beryllium peak erosion rate of the order of 0.1 nm sÀ1 These predictions are confirmed by extrapolation of experimental data from JET.180 This erosion rate would be acceptable from a component lifetime standpoint, especially during the low duty-factor operation of ITER However, the total amount of eroded material may be significant This material will most likely go to the divertor, and this will affect the composition of the divertor surface; therefore, it will affect the divertor performance and contribute to tritium codeposition and dust inventories Modeling of the influx of the eroded beryllium on the divertor is in progress to extrapolate from present machines and, in particular, to account for effects arising from material mixing including codeposition as expected in ITER Several studies have been recently published on this subject (see, e.g., Kirschner et al.207,208) 4.19.6.2.2.2 Erosion of the beryllium wall during ELMs Depending on the actual energy flux on the Be PFCs in ITER during ELMs, melt damage may or may not occur For Type I ELMs, which are compatible with the ITER divertor lifetime ($10 MJ ELMs16,18), the expected energy flux on the main chamber in ITER will be in the range of 2–3 MJ The area of the wall over which this flux will be distributed is $30–60 m2, for a toroidally symmetric energy deposition This leads to ELM energy fluxes $0.02–0.08 MJ mÀ2 on the main chamber wall, which will cause no Be melting at all If toroidal asymmetries and/or poloidal structures dominate the ELM energy deposition on the first wall, a substantial reduction of the first-wall effective area for energy deposition is expected (by a factor of $5) In this case, the ELM energy fluxes on the first wall would be 0.1–0.4 MJ mÀ2, which can cause up to 18 mm of 656 Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices melting, lasting $300 ms.209 Figure 21 shows the results of an analysis carried out with the code described in Raffray and Federici.210,211 The erosion lifetime, expressed in number of ELMs or corresponding ITER full power pulses (approximately 700 ELMs/pulses for a Be target initially 10 mm thick) is found to sharply decrease above a certain ELM energy threshold Depending on the duration of the ELM event, the threshold energy density varies between 0.2 and 0.7 MJ mÀ2 For comparison, the results of a W wall are also shown More recently, analysis has been carried out using more sophisticated modeling tools and the results are described elsewhere.212 From the JET Be divertor experience, we expect that only a very small part of the melt layer produced during each ELM will be mobilized (typically 95% of the plasma energy by radiation and substantially reduce mechanical stress on the vessel caused by poloidal halo currents Nevertheless, there remains some concern that even mitigated disruptions could damage the Be wall At some locations (mostly at the upper inboard region and the lower region of the first wall), the Be armored PFCs must withstand a certain number of ‘slow’ thermal transients resulting from loss-ofcontrol of plasma position during VDEs Typical parameters for these events are 60 MJ mÀ2 over 0.3 s In contrast to thermal quench disruptions, VDEs lead not only to significant erosion or melting, but also to high heat fluxes and a subsequent temperature increase at the armor/heat sink interfaces that can result in a failure of the armor/heat sink joints.217 As a matter of fact, because of their short duration (108 s vs >107 s), high duty cycle, and the higher temperature of the fluid to cool the PFCs to maintain a high plant energy conversion efficiency The higher surface temperature of the PFC will affect particle recycling, tritium uptake, chemical erosion, and material-mixing effects Erosion rates at the divertor target are very difficult to predict in conventional fusion power plant concepts with solid high-Z targets because the net erosion or deposition is strongly dependent on plasma parameters The fraction of ions arriving above the sputtering threshold is crucial, as is the efficiency of the prompt local redeposition ELMs have not really been considered in this context but we can see from the analysis in Section 4.19.6.2.2.2 that ELMs in power plant systems will have to be extremely small – much smaller than will be allowable in ITER, which still has a relatively low duty cycle The ideal would in fact be a quiescent ELM free high density steady state edge plasma Calculations of the minimum erosion rate for the main wall are somewhat more robust as there has to be a hot plasma in the main chamber and the rate of leakage of neutrals into the main chamber from the divertor can be calculated using Monte-Carlo codes In a recent study, Be, C, Fe, Cu, Mo, and W walls were compared,206 with the conclusion that in all cases the erosion rate was 1–2 t per year of continuous operation A Be or CFC wall will erode too rapidly in a reactor and the large amount of eroded material might give rise to deleterious problems as far as 659 control of the tritium and dust inventories are concerned A medium-Z material, such as Fe, does not seem to be acceptable purely from the standpoint of erosion lifetime As molybdenum is unfavorable for long-term activation problems, W is the best and only solution we have available for a reactor Effects of plasma contamination from Mo and W at the wall of tokamaks are being addressed in Alcator C-Mod, ASDEX-Upgrade and in the near-future at JET There is considerable gross erosion by sputtering for all materials The contributions of ions and neutrals from the plasma to this erosion are of the same order of magnitude The integrated total erosion due to ions and the energetic neutrals for the different wall materials (Figure 24) show that because of the larger sputtering yields for the low-Z materials, the number of atoms eroded for these materials is a factor of 10–20 larger than for high-Z materials such as W However, the total mass loss is similar for all materials, up to several kilograms per day or about t per year The maximum wall thinning for the low-Z materials is about 3.5 mm yearÀ1, while for high-Z materials, such as W, it is 0.22 mm yearÀ1, that is, lower by about a factor of 15 These values are in reasonable agreement with erosion measurements at the JET vessel walls.223 With respect to wall thinning, W is favorable for the use at the vessel walls because it has the longest ‘erosion lifetime’ (Figure 24(b)) With respect to plasma contamination, the probability of the eroded atoms entering into the plasma core, their lifetime in the plasma core, and the tolerable concentration of these ions in a burning fusion plasma all have to be taken into account.206 The tolerable concentration of W in the plasma is nearly three orders of magnitude lower than for low-Z atoms, such as Be and C However, recent observations have shown that W can be effectively removed from the plasma center by central heating.224 As this central heating is natural for burning plasmas, W may be a possible plasmafacing material, even from the viewpoint of plasma contamination The ion and neutral flux densities on the vessel walls are of the order of 1020 mÀ2 sÀ1, which may be critical with respect to the tritium implantation, accumulation in and permeation through the vessel walls 4.19.7 Concluding Remarks Beryllium is a low-density metal that is used in a number of industries, including the nuclear, automotive, aerospace, defense, medical, and electronics 660 Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices 25 (a) Time (year) to erode mm Wall erosion (atoms s-1) 1022 Be Ctot Cchem 1021 Cphys Cu Fe Mo W 1020 20 W 15 10 Mo Be C Cu 0 10 20 30 40 50 60 70 Atomic number of the material 80 (b) Fe 10 20 30 40 50 60 70 Atomic number of the material 80 Figure 24 (a) Integrated gross erosion due to ions and the energetic neutrals (b) Upper estimate for the time until a thickness of mm is eroded by sputtering at the area of largest erosion, that is, at the reference distance of about 11 m Reproduced with permission from Behrisch, R.; Federici, G.; Kukushkin, A.; Reiter, D J Nucl Mater 2003, 313–316, 388–392 industries, for various applications because it is exceptionally strong, is light in weight compared with other metals, has high heat-absorbing capability, and has dimensional stability in a wide range of temperatures Beryllium has been considered for many years as a primary candidate for protection of PFCs in tokamaks because it offers distinct advantages when compared with alternative materials such as carbon and tungsten It has a low atomic number and is an excellent oxygen getter The interaction of beryllium with tritium is also significantly weaker than that of carbon, leading to potentially reduced tritium inventory Beryllium does not form stable hydrides above 300  C, so there should be very little trapping expected in codeposited layers formed at such temperatures in the divertor after sputtering, although work is still underway to clarify this problem However, beryllium has a relatively high physical sputtering rate and a relatively low melting temperature and as such is more susceptible to melting damage that may occur in a tokamak during thermal transients In addition, because of its toxicity, special precautions are needed for working with beryllium, either for manufacturing or research investigation purposes Beryllium has been used with success in various tokamaks in the past mainly because of its ability to getter oxygen and to improve plasma performance In particular, its successful deployment in JET that started in 1989 and is continuing today with the installation of a completely new beryllium wall is the main rationale for the selection of beryllium as a plasma-facing material for the first wall of ITER, on the basis of a combination of plasma compatibility and design considerations This paper reviewed the properties of beryllium that are of primary relevance for plasma protection applications in magnetic fusion devices (i.e., PWIs, thermal and mechanical properties for power handling, fabricability and ease of joining, chemical reactivity, etc.) together with the available knowledge on performance and operation in existing fusion machines Special attention was given to beryllium’s erosion and deposition, formation of mixed materials, and its hydrogen retention and release characteristics These phenomena have a profound impact on component design, machine operation, and safety Extensive data on the behavior of Be with plasmas have been collected from existing tokamaks and simulators during the last two decades and this has enabled great strides to be made in our understanding of the PWI processes involved However, there are many issues for which there are still uncertainties and we will only learn from operating the next two major experiments that foresee the use of large amounts of Be ( JET and ITER) Much work remains to be done in this area and more machine operational time and diagnostics dedicated to PWIs are required Initiatives on these fronts, together with modeling of the results, are essential to advance the understanding of PWIs This includes (1) the possible surface damage (melting) during transients such as ELMs and disruptions and its implications for operations and (2) the problem of beryllium mixing with other armor materials and in particular the issue of codeposition of tritium with Be, which is eroded from the first wall and deposited at the divertor targets Such material may also be locally redeposited into shadowed areas of the shaped ITER first wall Both issues are part of Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices ongoing research, the initial results of which are being taken into account in the ITER design so that the influence of these two factors on ITER operation and mission are minimized For example, ITER will very likely employ, ELM control systems based on pellets and RMP coils, disruption mitigation systems, and increased temperature baking of the divertor to release T from Be-codeposits Dust generation is still a process which requires more attention Conversion from gross or net erosion to dust and the assessment of dust on hot surfaces need to be investigated At the time of writing this paper, the ITER first wall and shielding blanket is undergoing a major redesign effort to overcome some of the main shortcomings that were identified in the context of a recent design review scrutinizing the internal components Complex and interrelated materials, manufacturing, and design issues were briefly reviewed in this paper together with the progress of the manufacturing technologies being used and tested to demonstrate the durability of the joints A critical feature of the ITER first-wall design is the beryllium to copper alloy bond The joints must withstand the thermal, mechanical, and neutron loads and the cyclic mode of operation, and operate under vacuum, while providing an acceptable design for lifetime performance and reliability The availability of reliable joining technologies has a large impact on the design of the PFCs and on the overall cost of these components The status of the available techniques presently considered to join the Be armor to the heat sink material of Cu alloys for the fabrication of Be-clad actively cooled components for the ITER first wall was discussed During earlier ITER design phases, the feasibility of manufacturing reliable Be–CuCrZr joints was demonstrated The results of the performance and durability heat flux tests conducted in the framework of the further ITER first-wall qualification program were described This program has been launched and is in progress in the ITER parties in order to qualify the design and manufacturing routes The integrity of this bond must be assured for reliable ITER performance whatever process is used to fabricate joints The original procurement sharing that assigned the fabrication of first-wall panels up to six parties was seen as a risk and the number of parties supplying these critical components has now been reduced to three, Europe, the Russian Federation, and China The selection of specific grades of specific beryllium for the ITER first wall was described The 661 effects of neutron irradiation on the degradation of the properties of beryllium itself and on the joints were also analyzed Some of the changes are important while others are not significant for the ITER conditions Change of thermal conductivity and swelling are not important because of the low fluence The bulk tritium retention in neutron irradiated Be is expected to be significantly less than tritium retention in the codeposited layers The most critical consequence of neutron irradiation under ITER conditions is embrittlement This is typical of all grades of beryllium The structural integrity of neutron irradiated brittle Be is a key issue Embrittlement of neutron-irradiated Be could lead to increased thermal erosion and crack formation, which is also observed to occur for unirradiated beryllium under severe transient heat loads These cracks could serve as thermal fatigue crack initiation sites and accelerate this type of damage While this effect has not been extensively studied because of the difficulty of simulating disruptions in the laboratory, it may not be a critical issue as thermal fatigue cracks form after a few hundred cycles in most materials and they grow only to depths where the thermal stress level is above the yield stress On the basis of the information available from existing fusion machines, we discussed the problems that are still at issue in the design and operation of ITER This includes, in particular, the problem of erosion/ damage and the problem of up-take and control of tritium in the beryllium-based codeposited films Finally, on the basis of these results some tentative and speculative consideration of the limited prospects that beryllium has in future reactors was offered The 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et al J Nucl Mater 2005, 337–339, 852–856 ... a Plasma- Facing Material for Near- Term Fusion Devices 4. 19. 3.2 Hydrogen Retention and Release Characteristics 4. 19. 3.2.1 Implantation The use of beryllium as a plasma- facing material in tokamaks... compared to that ỵ calculated for a (0.25, 0 .47 , 0.28) mix of Dỵ, Dỵ , and D3 ; also shown is the measured yield from such a plasma Beryllium as a Plasma- Facing Material for Near- Term Fusion Devices. .. solutions for issues of plasma power handling and plasma wall interactions (PWIs) are required for the realization of a commercially attractive fusion reactor A mix of several plasma- facing materials