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

ITER Organization, St Paul Lez Durance, France

ß 2012 Fusion for Energy (F4E) Published by Elsevier Ltd All rights reserved.

4.19.2.2 Brief History of Plasma-Facing Materials in Fusion Devices 626

4.19.4.2 Selection of Beryllium Grades for ITER Applications 640

4.19.5.1 Joining Technologies and High Heat Flux Durability of the Be/Cu Joints 644

4.19.5.1.2 High heat flux durability of unirradiated Be/Cu joints 646

4.19.6 Tokamak PFC Design Issues and Predictions of Effects in ITER During Operation 650

4.19.6.2 Predictions of Effects on the ITER Beryllium Wall During Operation 653

621

4.19.6.2.1 Safety issues in ITER 653

4.19.6.3 Prospect of Using Beryllium in Beyond-ITER Fusion Reactors 659

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 ANFIBE Computer code for ANalysis of

Fusion Irradiated BEryllium ASDEX-

Upgrade

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.

ATC Adiabatic Toroidal Compressor

CFC Carbon-fiber composite

CIP Cold isostatic pressing

DIII-D A medium-sized tokamak, but the

largest tokamak still operational in the United States Operated by General Atomics in San Diego DIMES Divertor Material Evaluation

Studies, a retractable probe that allows the insertion and retraction

of test material samples to the

DIII-D divertor floor, for example, for erosion/deposition studies.

DS-Cu Dispersion-strengthened copper

EAST Experimental advanced

superconducting tokamak – an experimental superconducting

tokamak magnetic fusion energy reactor in Hefei, the capital city of Anhui Province, in eastern China ELMs Edge localized modes

FISPACT Inventory code included in the

European Activation System FZJ Forschungszentrum Juelich,

Germany HIP Hot isostatic pressing INEEL Idaho National Engineering and

Environmental Laboratory Now Idaho National Laboratory (INL) ISX Impurity study experiment (ISX-A

and ISX-B where two tokamaks operated at Oak Ridge National Laboratory)

ITER 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 JET 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 1 for DT operation (Q ¼ 0:95

in 1997)

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

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

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

4.19.1 Introduction

Beryllium, once called ‘the wonder metal of the future,’1

is a low-density metal that gained early prominence

as a neutron reflector in weapons and fission researchreactors It then found a wide range of applications

in the automotive, aerospace, defense, medical, andelectronic industries Also, because of its uniquephysical properties, and especially favorable plasmacompatibility, it was considered and used in the pastfor protection of internal components in variousmagnetic fusion devices (e.g., UNITOR, ISX-B, JET).Most important future (near-term) applications in thisfield include (1) the installation of a completely newberyllium wall in the JET tokamak, which has beencompleted by mid of 2011 and consists of1700 solid

Be tiles machined from 4 t of beryllium; and (2) ITER,the world’s largest experimental facility to demon-strate the scientific and technical feasibility of fusionpower, which is being built in Cadarache in the South

of France ITER will use beryllium to clad the firstwall (700 m2

for a total weight of about 12 t of Be).Although beryllium has been considered for otherapplications in fusion (e.g., as neutron multiplier inthe design of some types of thermonuclear breedingblankets of future fusion reactors and for hohlraums

in inertial confinement fusion), this chapter will

be limited to discussing the use of beryllium as aplasma-facing material in magnetic confinementdevices, and in particular in the design, research,and development work currently underway forthe JET and the ITER tokamaks Considerationsrelated 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 thermonuclearplasma and the surrounding solid material environ-ment has been arguably one of the greatest technicalchallenges of ITER and will continue to be a chal-lenge for the development of future fusion powerreactors The interaction between the edge plasmaand the surrounding surfaces profoundly influencesconditions in the core plasma and can damage thesurrounding material structures and lead to longmachine downtimes for repair Robust solutions forissues of plasma power handling and plasma–wallinteractions (PWIs) are required for the realization

of a commercially attractive fusion reactor A mix

of several plasma-facing materials is currently posed in ITER to optimize the requirements of areaswith different power and particle flux characteristics

pro-(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

for-mation of material mixtures, whose behavior is still

uncertain and requires further investigation

The use of beryllium for

plasma-facing-component (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

beryl-lium that were held in the past, bringing together

leading researchers in the field of beryllium

tech-nology and disseminating information on recent

progress in the field.4 Comprehensive reviews have

also appeared recently in specialized journals5,6

con-taining state-of-the-art information on a number of

topics such as manufacturing and development of

coat-ing 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,

fab-ricability and ease of joining, chemical reactivity, etc.)

together with the available knowledge on

perfor-mance 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,

com-ponent 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 erties such as erosion/deposition, hydrogen retentionand release, and chemical effects such as materialmixing, all of which influence the selection of beryl-lium as armor material for PFCs Section 4.19.4

prop-briefly reviews a limited number of selected sical and mechanical properties of relevance forthe fabrication of heat exhaust components and theeffects of neutron irradiation on material properties

phy-Section 4.19.5 describes the fabrication issues andthe progress of joining technology and high heat fluxdurability of beryllium-clad PFCs Section 4.19.6

describes the main issues associated with the JETand ITER first-wall designs and discusses some con-straints foreseen during operation The prospects ofberyllium for applications in fusion reactors beyondITER are briefly discussed Finally, a summary isprovided inSection 4.19.7

A detailed discussion on this subject is beyond thescope of this review The relevant PWIs are compre-hensively reviewed by Federici et al.7,8More recentinterpretations of the underlying phenomena andimpact on the ITER device can be found in Roth

et al.9Here we summarize some of the main points.PWIs critically affect tokamak operation in manyways Erosion by the plasma determines the lifetime

of PFCs, and creates a source of impurities, which cooland dilute the plasma Deposition of material ontoPFCs alters their surface composition and, depending

on the material used, can lead to long-term tion of large in-vessel tritium inventories This latterphenomenon is especially exacerbated for carbon-based materials but there are still some concerns withberyllium Retention and recycling of hydrogen fromPFCs affects fuelling efficiency, plasma density control,and the density of neutral hydrogen in the plasmaboundary, which impacts particle and energy transport.The primary driver for the interactions betweenthe core plasma, edge plasma, and wall is the powergenerated in the plasma core that must be handled bythe surrounding structures Fusion power is obtained

accumula-by the reaction of two hydrogen isotopes, deuterium(D) and tritium (T), producing an a-particle and afast neutron Although the kinetic energy carried bythe 14.1 MeV neutron escapes the plasma and could

be converted in future reactors beyond ITER tothermal energy in a surrounding blanket system, the

kinetic energy of thea-particle is deposited in the

plasma The fraction of this power that is not radiated

from the plasma core as bremsstrahlung or line

radi-ation (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

represen-tation of the regions of the plasma and boundary

walls in a divertor tokamak is portrayed inFigure 1

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

surface’ (LCFS) or ‘separatrix.’ Magnetic field surfacesinside the LCFS are closed, confining the plasma ions.The edge region, just inside the LCFS, contains signif-icant levels of impurities not fully ionized, and perhapsneutral particles Impurity line radiation and neutralparticles transport some power from here to the wall.The remaining power enters the region outside theLCFS either by conduction or, depending on thedegree to which neutrals penetrate the plasma, byconvection This region is known as the scrape-off-layer (SOL) as here power is rapidly ‘scraped off ’ byelectron heat conduction along open field lines, whichare diverted to intersect with target regions that areknown as ‘divertors.’ Poloidal divertors have been verysuccessful at localizing the interactions of plasma ionswith the target plate material in a part of the machinegeometrically distant from the main plasma where anyimpurities released are well screened from the mainplasma and return to the target plate

The plasma density and temperature determinethe flux density and energy of plasma ions strikingthe plasma-wetted surfaces These, in turn, deter-mine the rate of physical sputtering, chemical sput-tering, ion implantation, and impurity generation.The interaction of the edge plasma with the sur-rounding solid material surfaces is most intense inthe vicinity of the ‘strike point’ where the separatrixintersects the divertor target plate (see inset in

Figure 1) In addition, the plasma conditions mine where eroded material is redeposited, and

deter-to what degree codeposition of tritium occurs at thewall The plasma power flow also determines the level

of active structural cooling required

Typical plasma loads and the effects expectedduring normal operation and off-normal operation

in ITER are summarized inTable 1.Because of the very demanding power handlingrequirements (predicted peak value of the heat flux inthe divertor near the strike-points is >10 MW m2)and the predicted short lifetime due to sputteringerosion arising from very intense particle fluxes(1023

–1024 particles m2s1) and damage duringtransient events, beryllium has been excluded fromuse in the ITER divertor and is instead the materialselected for the main chamber wall of ITER.Recent observations in present divertor tokamakshave shown that plasma fluxes to the main wallare dominated by intermittent events leading to fastplasma particle transport that reaches the PFCs alongthe magnetic field (see Loarte et al.10and referencestherein) The quasistationary heat fluxes to the mainwall are thought to be dominated by convective

Magnetic flux surfaces

First wall Separatrix (LCFS)

Separatrix (LCFS) X-point

Plasma core

Baffle

Vertical divertor target plate Private flux

region Separatrix strike point

Pump Divertor region

Edge region Scrape-off layer

Figure 1 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.

transport,11but 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 m2 in the ITER QDT¼ 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.13–15There are still large

uncer-tainties 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 m2 (controlledELMs) to 20 MJ m2(uncontrolled ELMs).16,17Evenfor controlled ELMs, such energy fluxes are likely tocause melting of up to several tens of micrometers ofberyllium at the exposed edges,18which could causeundesirable impurity influxes at every ELM.10,11

4.19.2.2 Brief History of Plasma-FacingMaterials in Fusion Devices

PWIs have been recognized to be a key issue inthe realization of practical fusion power since thebeginning of magnetic fusion research By the time

of the first tokamaks in the 1960s in the USSR andsubsequently elsewhere, means of reducing the level ofcarbon and oxygen were being employed.19,20Theseincluded the use of stainless steel vacuum vessels andall-metal seals, vessel baking, and discharge cleaning.Ultimately, these improvements, along with improvedplasma confinement, led to the first production ofrelatively hot and dense plasmas in the T3 tokamak(1 keVand 3  1019

m3).21,22These plasmas, whilebeing cleaner and with low-Z elements fully stripped inthe core, still had unacceptable levels of carbon, oxy-gen, and metallic impurities The metallic contamina-tion inevitably consisted of wall and limiter materials

Table 1 Major issues associated with operation of ITER PFCs

Radiation and particle heat CFCa Chemical erosion evaporation

brittle destruction and tritium codeposition

Erosion lifetime and component replacement

Large particle fluxes

and safety

ELM’s

Slow-high power

transients Divertor –

Moderate power transients

First wall Plasma contact during

aW is also considered as an alternative.

bMultifaceted asymmetric radiation from the edge (MARFE).

cVertical displacement event (VDE).

Early in magnetic fusion research, it was

recog-nized 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

posi-tioned to ensure at least several centimeters gap

between the plasma edge and more delicate

struc-tures 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)27and Princeton Large Torus (PLT),28

and titanium in poloidal divertor experiment (PDX).29

Poloidal divertors have been very successful at

loca-lizing 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.30By 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’

sur-face, 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

radia-tion In addition, the surface does not melt if

over-heated – 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

By the mid-1980s, many tokamaks were operatingwith graphite limiters and/or divertor plates Inaddition, extensive laboratory tests and simulations

on graphite had begun, primarily aimed at standing the chemical reactivity of graphite withhydrogenic plasmas, that is, chemical erosion Earlylaboratory results suggested that carbon would beeroded by hydrogenic ions with a chemical erosionyield of Y  0.1 C/Dþ, a yield several times higherthan the maximum physical sputtering yield Anotherprocess, radiation-enhanced sublimation (RES), wasdiscovered at elevated temperatures, which furthersuggested high erosion rates for carbon Carbon’s abil-ity to trap hydrogenic species in codeposited layerswas recognized These problems, along with graphite’spoor mechanical properties in a neutron environment(which had previously been known for many yearsfrom fission research33), led to the consideration ofberyllium as a plasma-facing material This was pri-marily promoted at JET.34A description of the opera-tion experience to date with Be in tokamak devices isprovided inSection 4.19.2.3

under-At present, among divertor tokamaks, carbon is thedominant material only in DIII-D Alcator C-Mod

at Massachusetts Institute of Technology (MIT),USA35 uses molybdenum ASDEX-Upgrade (AxiallySymmetric Divertor Experiment) is fully clad withtungsten,36 and JET has completed in 2011 a largeenhancement programme37that includes the installa-tion of a beryllium wall and a tungsten divertor Newsuperconducting tokamaks, such as Korea Supercon-ducting Tokamak Advanced Reactor (KSTAR) inKorea38and experimental advanced superconductingtokamak (EAST) in China,39employ carbon as materialfor the in-vessel components, but with provisions toexchange 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 onplasma performance and operation, (2) maximumoperational flexibility at the start of operation, and(3) minimum fuel retention for operation in the DTphase This compromise is met by a choice of threeplasma-facing materials at the beginning of opera-tions (Be, C, and W) It is planned to reduce thechoices to two (Be and W) before DT operations

in order to avoid long-term tritium retention in bon codeposits during the burning plasma phase.Beryllium has been chosen for the first-wall PFCs

car-to minimize fuel dilution caused by impuritiesreleased from these surfaces, which are expected to

have the largest contamination efficiency.40–44

More-over, the consequences of beryllium contamination

on fusion performance and plasma operations are

relatively mild This has been demonstrated by

experiments in tokamaks (seeSection 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

rede-posited 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

temper-ature 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

mitiga-tion 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

experi-ments were performed by UNITOR,45 which were

then followed by ISX-B.46 Both tokamaks

investi-gated 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 Althoughthe concentration of beryllium in the plasma wasincreased, its contribution to Zeff (the ion effectivecharge of the plasma Zeff provides a measure forimpurity concentration) was more than compensated

by the reduction of oxygen, carbon, and metal rities.45The plasma Zeffwas observed to be reducedfrom 2.4 to near unity with beryllium It must be notedthat there was a negative aspect associated with beryl-lium operation during the ISX-B campaign Thereduction in plasma impurities was not observeduntil the limiter surface was partially melted causingberyllium to be evaporated and coated on the firstwall Once melting did occur, the droplets madesubsequent evaporation more likely but hard to con-trol The consequent strong reduction in plasmaimpurities associated with gettering then made dis-charge reproducibility hard to obtain However, if amuch larger plasma contact area is already coveredwith Be, one does not need to rely on limiter melting

impu-to obtain the beneficial effect of beryllium This effectcould be achieved by using large area beryllium lim-iter, or coating the inside wall with beryllium whichwas the approach taken by JET when it introducedberyllium in 1989

Large tokamak devices such as JET had found

it very difficult to control the plasma densitywith graphite walls as the discharge pulse length gotlonger Motivated by the frequent occurrence of

a phenomenon that plagued the earlier campaigns –the so-called carbon blooms due to the overheating ofpoorly designed divertor tiles and the subsequentinflux of carbon impurities in the plasma due

to evaporation – JET decided to use beryllium as aplasma-facing material

Thin evaporated beryllium layers on the graphitewalls were used initially (100 A˚ average thicknessper deposition) on the plasma-facing surface of thedevice Subsequently, beryllium tiles were installed

on the toroidal belt limiter

The main experimental results with beryllium can

be summarized as follows:

1 The concentration of carbon and oxygen in theplasma were 4–7% and 0.5–2%, respectively,when graphite was used as belt limiter With aberyllium belt limiter, the carbon content wasreduced to 0.5% and oxygen became negligible,because of oxygen gettering by beryllium Duringohmically heated discharges, the concentration

of beryllium remained negligible even thoughberyllium was the dominant impurity

2 While the value of Zeffwas3 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

3 The fuel density control was greatly improved

with the beryllium limiter and beryllium

evapo-rated wall Gas puffing to maintain a given plasma

density was typically 10 times larger when using

beryllium than graphite

Following the beryllium limiter experience,

diver-tor beryllium targets were installed in JET for two

configurations An extensive set of experiments with

toroidally continuous X-point divertor plates was

car-ried out in JET in the period 1990–1996 to characterize

beryllium from the point of view of its

thermomecha-nical 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

pro-gressive 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

tem-perature reached 1300C 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

low-density H-modes, despite the large scale distortion of

the target surface caused by the melt layer

displace-ment and splashing due to the previous 25 high

power discharges48,51(seeFigure 252

) This strated that it was possible to use the damaged Be

demon-divertor target as the main power handling PFC if the

average power load was decreased, either by ing plasma density and radiative losses, or by strikepoint sweeping The damage did not prohibitsubsequent plasma operation in JET, but would seri-ously limit the lifetime of Be PFCs in long-pulseITER-like devices

increas-The latest results of the operation of JETwith beryllium have been reviewed recently byLoarte 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’ lium surfaces at different temperatures Retention ofplasma fuel species in both bulk and codepositedlayers of beryllium is then described As berylliumwill not be used as the exclusive plasma-facing mate-rial in future confinement devices, issues associatedwith mixed, beryllium-containing surfaces are alsoaddressed

beryl-4.19.3.1 Beryllium Erosion PropertiesThe term erosion is used to describe a group ofprocesses that remove material from a surface sub-jected to energetic particle bombardment Includedunder the general classification of erosion are pro-cesses such as physical sputtering, chemically assistedphysical sputtering, chemical sputtering, and thermallyactivated release from surfaces Of these processes,only chemical sputtering, where volatile molecularspecies are formed on the surface, appears to beinactive in beryllium

4.19.3.1.1 Physical sputtering of berylliumPhysical sputtering results from the elastic transfer

of energy from incoming projectiles to atoms on thesurface of the target material Target atoms can besputtered when the energy they receive from thecollisional cascade of the projectile exceeds the bind-ing energy of the atom to the surface The physicalsputtering rate is usually referred to as the sputteringyield, Y, which is defined as the ratio of the number ofatoms lost from a surface to the number of incidentenergetic particles striking the surface The lowerthe binding energy of surface atoms, the larger thephysical sputtering yield As physical sputtering can

be approximated using a series of binary collisionswithin the surface, it is relatively easy to estimate

Figure 2 Melting of the Joint European Torus Mk

I beryllium target plate tiles after plasma operation Image

courtesy of EFDA-JET.

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 targets54including 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.55It has also been

shown that it is possible to deplete the beryllium

surface of oxide by heating the sample to

tempera-tures exceeding 500C, where the beryllium below

the oxide can diffuse through the oxide to the

surface,56thereby allowing measurements on a clean

beryllium surface The comparison between the

calcu-lated 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

elec-trostatic 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

ions (Dþ, D2 þ, D3 þ) must be taken into account

as the incident molecule disassociates on impact

with the surface and a D2 þ ion becomes equivalent

to the bombardment of two deuterium particles with

one-half the incident energy of the original D2 þion

Figure 3shows the change to the calculated

sputter-ing yield when one includes the effects of molecular

ions in a plasma, compared to the calculated

sputter-ing yield from pure Dþion 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

back-ground 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 incidentflux and a new surface exhibiting unique character-istics may result

Nevertheless, the physical sputtering yield fromberyllium surfaces exposed to plasma ion bombard-ment has been measured in several devices Unfortu-nately, there is little consensus on the correct value ofthe physical sputtering yield In JET, the largestconfinement device to ever employ beryllium as aPFC sputtering yield measurements range fromvalues far exceeding47to values less58than one wouldexpect from the predictions of TRIM In the PlasmaInteraction with Surface Components ExperimentalStation B (PISCES-B) device, systematic experiments

to measure the physical sputtering yield routinelyshow values less59–61 than those expected fromTRIM This difference is shown in Figure 3, wherethe energy dependence of the calculated yield is com-pared to experimental measurements

Another primary difference between the tions in an ion beam device and those encountered

condi-in a plasma device has to do with the neutral densitynear the surface being investigated In an ion beamexperiment, the background pressure is kept very low

Measured yield in D plasma

Ion energy (eV)

0 10-5

Figure 3 Calculated sputtering yields from pure Dþbombardment at normal incidence compared to that calculated for a (0.25, 0.47, 0.28) mix of Dþ, D 2 þ , and D 3 þ ; also shown is the measured yield from such a plasma.

so that the surface being probed maintains its clean

properties On the other hand, the incident flux in a

plasma device is usually several orders of magnitude

larger than in an ion beam device, ensuring that the

surface remains clean because of the large incident

flux However, this plasma-facing surface undergoes

not only energetic ion bombardment, but also

bom-bardment by neutral atoms and molecules

The neutral density in plasma generators is

typi-cally on the order of 1020m3(a few millitorr) which

is necessary for breakdown of the plasma The

esti-mated neutral atom flux is approximately equal to the

incident ion flux to the surface61and it is often not

possible to alter significantly this flux ratio In the

case of a beryllium surface which can form a hydride

(see Section 4.19.3.1.3), the presence of adsorbed

deuterium on the surface could affect the measured

sputtering yield by decreasing the beryllium

concen-tration at the surface and altering the binding energy

of surface beryllium atoms

Some evidence of this effect may be discerned

in data from JET measurements of the beryllium

sputtering yield Two sets of sputtering yield

mea-surements have been reported from JET; one from

beryllium divertor plate measurements and the other

from beryllium limiter measurements In the divertor

region, one expects a neutral density similar to that

encountered in plasma generators (1020m3or more)

and the measured sputtering yield is lower than that

predicted by TRIM calculations.58When sputtering

measurements are made on the limiter, where the

neutral density is typically lower, the sputtering

yield agrees with, or exceeds, the calculated value.47

Of course, other issues such as impurity layers on the

divertor plate and angle of incidence questions tend to

confuse the results However, the data sets from JET

are consistent with the impact of neutral absorption

on the beryllium plasma-facing surface

Effects associated with plasma operation will need

to be taken into account when predicting sputtering

yields from different areas of confinement devices

In addition to the low-energy neutral atom flux and

higher-energy charge exchange neutral flux, the

impact of small impurity concentrations in the

inci-dent plasma flux will also have a large impact on the

expected sputtering yield Some of the implications

of the formation of a mixed-material surface are

discussed in the next section and inSection 4.19.3.3

4.19.3.1.2 Mixed-material erosion

As was pointed out in the previous section, it is

impor-tant to have accurate knowledge of a target’s surface

composition to predict its erosion rate A small rity concentration contained within the incidentplasma can drastically alter the surface composition

impu-of a target subjected to bombardment by the impureplasma Oxygen impurities in the plasma, either fromionization of the residual gas, or due to erosion fromsome other surface, will readily lead to the formation ofberyllium oxide on the surface of a beryllium target.Depending on the arrival rate of oxygen to the surfacecompared to the erosion rate of oxygen off the surface,one can end up measuring the sputtering rate of a cleanberyllium surface or a beryllium oxide surface Carefulcontrol of the residual gas pressure in ion beam sput-tering experiments55 has documented this effect.Unfortunately, it is not always so easy to control theimpurity content of an incident plasma

In the case of a magnetic confinement devicecomposed of groups of different plasma-facing mate-rial surfaces, erosion from a surface in one location ofthe device can result in the transport of impurities

to other surfaces throughout the device material surfaces are the result To first order, amixed-material surface will affect the sputtering ofthe original surface material in two ways The first israther straightforward, and is true even for materialswhich do not form chemical bonds, in that the surfaceconcentration of the original material is reducedthereby reducing its sputtering rate The second effectchanging the sputtering from the surface results fromchanges in the chemical bonding on the surface, whichcan either increase, or decrease the binding energy

Mixed-of the original material If the chemical bonds increasethe binding energy, the sputtering rate will decrease

If the bonding acts to reduce the surface bindingenergy, the sputtering rate will increase (assuming thechange in surface concentration does not dominate thiseffect) A recent review of mixed materials62providessome background information on the fundamentalaspects of general mixed-material behavior

If a plasma incident on a beryllium target containssufficient condensable, nonrecycling impurities (such

as carbon), it will affect the sputtering rate of theberyllium This effect was first referred to as ‘carbonpoisoning.’5,9,63 A simple particle balance modelhas been used to adequately explain the results forformation of mixed carbon-containing layers onberyllium at low surface temperature.64However, asthe target temperature increases, additional chemicaleffects, such as carbide formation, have to beincluded in the model

An interesting change occurs when the ing species is a mixture of carbon and oxygen

bombard-Measurements of the chemical composition of a

beryllium surface bombarded with a COþion beam

showed almost exclusive bonding of the oxygen

to the beryllium in the implantation zone.65,66

The formation of BeO on the surface left the carbon

atoms easily chemically eroded The amount of

oxy-gen present in the incident particle flux plays a strong

role in the final chemical state of the surface atoms

and their erosion behavior

The inverse experiment, beryllium-containing

plasma incident on a carbon surface, has also been

investigated.67–69In the case of beryllium impurities

in the plasma, a much more accurate measurement of

the impurity concentration was possible Contrary

to the carbon in beryllium experiments, a simple

particle balance model could not account for the

amount of beryllium remaining on the surface after

the plasma exposure Clearly, the inclusion of

chemi-cal effects on the surface needs to be taken into

account to interpret the results

Beryllium carbide (Be2C) was observed to form on

the surface of carbon samples exposed to

beryllium-containing deuterium plasma even during

bombard-ment at low surface temperature Carbide formation

will also act to increase the binding of beryllium

atoms to the surface and decrease the binding of

carbon atoms This effect will result in an increase

in the concentration of beryllium on the surface

compared to a simple particle balance equation and

must be included to understand the evolution of the

surface In addition, the formation of the carbide was

correlated with the decrease of carbon chemical

ero-sion70(seeSection 4.19.3.3.1 for more discussion of

the chemical erosion of the beryllium–carbon system)

4.19.3.1.3 Chemically assisted sputtering

of beryllium

The term chemically assisted physical sputtering

refers to the transfer of energy from an incident

particle to a molecule on the surface The energy

gained is sufficient to break any remaining bonds of

the molecule to other atoms on the surface resulting

in the release of the molecule, or a fragment of the

molecule, from the surface In the case of beryllium

bombarded by deuterium plasma, the sputtering of

beryllium deuteride was first recorded in JET71

dur-ing operation with a beryllium divertor plate Since

that time, a series of systematic investigations were

performed in PISCES-B to quantify the magnitude of

this erosion term.72,73

The results from PISCES-B show a surface

tem-perature dependence of the sputtering rate72of BeD

molecules The maximum in the BeD sputtering rate(at175C) corresponds with the onset of thermal

decomposition of BeD2molecules73from a dized sample of BeD2powder Even at the maximumloss rate, the chemical sputtering remains smallerthan the physical sputtering rate of beryllium atomsfrom the surface over the incident energy rangeexamined (50–100 eV) Molecular dynamics simula-tions have predicted,74 and subsequent measure-ments have validated the prediction, that chemicalsputtering can dominate physical sputtering of beryl-lium as the incident deuterium ion energy decreasesbelow 50 eV

standar-A distinction should be made between chemicalsputtering and chemically assisted physical sputtering.Chemical sputtering involves the formation and loss ofvolatile molecules from a surface In the case of beryl-lium deuteride, the molecule decomposes into a deu-terium molecule and a beryllium atom before itbecomes volatile, so at least to date there is no evidencefor chemical sputtering of beryllium during deuteriumparticle bombardment Documentation of the chemi-cally assisted physical sputtering of beryllium may beimportant for determining material migration patterns

in confinement devices and the identification of lium deuteride molecular formation in plasma-exposed surfaces may also help explain the hydrogenicretention properties of beryllium

beryl-4.19.3.1.4 Enhanced erosion at elevatedtemperatures

In addition to the temperature-dependent chemicalsputtering of beryllium when exposed to deuteriumplasma, another temperature-dependent loss term ispresent in beryllium exposed to plasma bombard-ment at elevated temperature The classical picture

of the temperature dependence of erosion fromchemically inert surfaces exposed to energetic parti-cle bombardment is composed of the superposition

of a constant physical sputtering yield with an nentially varying thermal sublimation curve Theclassical picture is contradicted, however, by experi-ments that show an exponential increase in erosion

expo-at lower temperexpo-ature thexpo-at cannot be explained byclassical thermodynamic sublimation First observed

by Nelson75 for a variety of metal surfaces, similarresults have been measured for Be,76,77 W,78 and

C79,80surfaces In the case of carbon, this mechanismhas been called RES

In the case of beryllium, two explanations havebeen proposed and both rely on the large flux of ionsincident during plasma bombardment to modify the

plasma-facing material surface In the first, the

inci-dent plasma ions, in addition to creating sputtered

atoms from the surface, also create a population of

surface adatoms An adatom is an atom from a lattice

site on the surface that has gained sufficient energy to

leave its lattice location, yet does not have sufficient

energy to escape from the surface as a sputtered

atom The atom then occupies a site on top of the

regular lattice sites Because an adatom does not have

the same number of adjacent atoms as those in the

lattice, it is less strongly bound to the surface and can

therefore sublime at a lower temperature than one

associates with equilibrium thermodynamic

sublima-tion In the second explanation, incident plasma ions

that have thermalized somewhere below the surface

of the lattice exert a stress on the surface atoms of the

target again resulting in a lower binding energy of

the surface atoms to the bulk of the material

Measurements show atoms are being lost from the

surface at thermal energies,77rather than the energy

associated with sputtered particles (i.e., on the order

of electron volts) This seems to verify the loss

mech-anism that occurs because of the thermal release of an

ensemble of particles with a lower surface binding

energy than that of bulk atoms of that element

Addi-tional measurements at elevated temperature have

documented the variation in Be atom surface loss

rate with changes of the incident flux of energetic

particles.81 The larger the incident flux, the lower

the onset temperature for the enhanced erosion

The implication of this enhanced loss rate at

ele-vated temperature is a reduction of the permissible

operating temperature of any plasma-facing material,

or alternatively that the lifetime of a component

operating at extreme temperature may be less than

that expected based on the predictions from classical

surface loss terms

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 has prompted many experimental studies

of retention and emission of hydrogen implanted into

beryllium-like metals from ion-beams or plasmas

References and discussions of these studies can be

found in reviews.82–85Here, we review those studies

which are relevant to H retention in Be in a fusion

plasma environment This section is mainly excerpted

from Federici et al.7Two basic parameters for

under-standing H retention are the hydrogen diffusivity and

solubility Studies of solubility and diffusivity arereviewed in Causey and Venhaus85and Serra et al.86

Figures 487–90

and587,91,92

show experimental valuesfor hydrogen solubility and diffusivity in W and Be.For Be there are significant differences betweenresults from various studies These differences may

be due to effects of traps and surface oxide layers.The presence of bulk traps tends to increase themeasured values of solubility and to decrease the mea-sured values of diffusivity (see Federici et al.7),especially under conditions where the concentration

1

2

3 Be

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.

W

1a

1b 2

3 Be

1 ´10 -9 1´10 -8

of hydrogen in solution is smaller than the

concen-tration of traps For this reason, studies done on

materials of higher purity and crystalline perfection,

and at higher temperatures and with higher

concen-trations of hydrogen in solution, tend to give more

reliable results The porosity and oxide inclusions

present in beryllium produced by powder metallurgy

are also likely to lead to inconsistent results in

mea-surements of hydrogen solubility and diffusivity

In the Be experiments, the effects of traps were not

characterized and may be dominant One firm

con-clusion is that the solubility of hydrogen is very low

in both Be and W

Many studies have been done on the retention and

emission of H implanted into materials to provide

data needed to predict H retention in fusion reactor

environments.Figure 6shows the retention of 1 keV

deuterium implanted into Be at 300 K versus

inci-dent fluence, measured by thermal desorption.93

D retention in Be was close to 100% at low fluences

but saturated at high fluences Earlier nuclear

reac-tion analysis (NRA) measurements of D retained in

Be within 1 mm of the surface gave very similar

results.94 This saturation behavior indicates that

D implanted into Be at 300 K does not diffuse, but

accumulates until it reaches a limiting concentration

of 0.3–0.4 D/Be within the implantation zone

At high fluences, the implanted zone becomes porous

allowing additional implanted D to escape This

saturation mechanism is confirmed by electron copy, which shows bubbles and porosity in the implan-tation zone after high fluence H implantation.95Saturation of retention by the same mechanism isobserved for D implanted into stainless steel at

micros-150 K where the D is not mobile.96H retention in Beincreases with increasing ion energy and decreaseswith increasing sample temperature.84,97The retention

of 1 keV deuterium implanted into W and Mo at

300 K98is also shown for comparison inFigure 6

of California, San Diego-Plasma Interaction withSurface Components Experimental Station B (UCSD-PISCES-B)), a tokamak divertor plasma (DIII-D-DIMES), and a neutral beam (NB-JET) In some

of these studies carbon deposition or formation ofcarbide or oxide surface layers occurred, which islikely to affect D retention The figure shows the

D retention in Be observed under a wide range ofexposure conditions The high fluence saturated con-centration tends to be lower at higher temperatures

2 3

It must be noted that this phenomenon is very

important because it implies that tritium inventories

and permeation due to implantation in beryllium

for ITER PFC applications should be significantly

lower than was previously estimated using classical

recombination-limited release at the plasma surface

A first attempt to model this saturation by allowing

the recombination coefficient to become

exponen-tially large as the mobile atom concentration near

the plasma-facing surface approaches a critical

value was made by Longhurst et al.99For Be,

calcula-tions suggest that the critical concentration is related

to the yield strength using Sieverts’ law of solubility

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 such as ITER, because

of implantation, diffusion, trapping, and

neutron-induced transmutation, will be of the order of 100 g

rather than the kilogram quantities estimated

previ-ously,100,101and most of that will result from

neutron-induced transmutations in the Be itself and from

trapping in neutron-induced traps Current

predic-tions of tritium inventory in ITER are briefly

dis-cussed inSection 4.19.6.2.1

Fusion neutrons will create vacancies and

intersti-tials in plasma-facing materials For metals at reactor

wall temperatures, these defects will be mobile and

will annihilate at sinks (e.g., surfaces or grain

bound-aries), recombine, or agglomerate into defect clusters

Vacancy agglomeration may also lead to the formation

of voids In beryllium, neutron-induced nuclear tions produce helium and tritium, which may betrapped at defects or precipitate as gas bubbles.These defects, resulting from neutron irradiation,will increase the retention of hydrogen, by increasingthe concentration of sites where diffusing hydrogencan precipitate as gas or become trapped as atoms.The effect of neutron irradiation on hydrogen reten-tion in metals is complex, but, in principle, this can bemodeled, provided the material parameters areknown, such as hydrogen diffusivity, solubility, trapbinding energy, and defect microstructure produced

reac-by the neutron irradiation For many metals, most ofthese parameters are known well enough to attemptsuch modeling For beryllium, however, uncertainties

in solubility, diffusivity, and trapping of hydrogenmake such modeling of hydrogen retention difficult.The problem of production of helium and tritium

by nuclear transmutation in beryllium itself is cussed inSection 4.19.4.4.5

dis-4.19.3.2.2 Beryllium codeposition

As deuterium retention in plasma-exposed berylliumtargets saturates after a given ion fluence (seeSection4.19.3.2.1), it is apparent that retention in codepositswill eventually be the dominant accumulation mech-anism with respect to beryllium PFCs This is pri-marily due to the fact that the thickness of acodeposit will continue to grow linearly with time

It is, therefore, critical to understand both the

250 200200

250

250 40 200 100 100

500

500 500500

500 500

500 300

300 200 200

540

700

Figure 7 Retention of deuterium and tritium in Be as a function of incident particle fluence For purposes of comparing results from different experiments using different ion energies, the data have been scaled to correspond to an equivalent

100 eV deuterium ion energy Numerical values next to the symbols and in the legend are specimen exposure temperatures,

in degrees Celsius Reproduced with permission from Anderl, R A.; et al J Nucl Mater 1999, 273, 1–26.

retention amounts and the release behavior of

hydro-gen isotopes from beryllium codeposits In this

sec-tion, a ‘codeposit’ includes both the codeposition

(where a BeD or BeD2 molecule is deposited on a

surface) and co-implantation (where deposited layers

of beryllium are bombarded with energetic hydrogen

isotopes) processes

Initial interpretation of studies of beryllium

code-posits were made difficult by relatively high oxygen

impurity content within the codepositing surface.102,103

Subsequent measurements104with lower oxygen

con-tent seemed to indicate that the oxygen level within

the codeposit was correlated to the level of hydrogen

isotope retention in the codeposit The other variable

that was identified to impact the retention level in

these studies was the temperature of the codepositing

surface

Measurements seriously questioning the

impor-tance of oxygen on the retention level in beryllium

codeposits were made by Baldwin et al.105In this data

set, the oxygen content throughout the codeposit was

measured by depth profiled X-ray photoelectron

spectroscopy and the oxygen content did not

corre-late with the deuterium retention level (Figure 8),

although the temperature of the codepositing surface

was still a dominating term in determining the

deu-terium retention level Later, more detailed

measure-ments confirmed that the presence of a beryllium

oxide surface layer was not correlated with anincrease in retention in beryllium.106

A systematic study of beryllium codeposition lowed,107identifying three experimental parametersthat seemed to impact the retention level in a code-posit Along with the surface temperature, the inci-dent deuterium energy and the beryllium depositionrate were determined to be influential scaling para-meters The previously reported data in the literaturewas also evaluated using the derived scaling andfound to agree with the predictions of the retentionlevels measured under the various experimentalconditions present in the different machines Laterthe derived scaling was revised108to use the ratio ofthe fluxes of the codepositing species, rather thanthe deposition rate to permit more accurate extrapo-lation to conditions expected in the edge of confine-ment devices

fol-The ability to predict the level of tritium retention

in beryllium codeposits is an important aspect of

a safety program; however, developing techniques

to remove the trapped tritium from codeposits is

a more important issue The deuterium releasebehavior during thermal heating of beryllium code-posits has been investigated.109The results show thatthe maximum temperature achieved during a bake-out is the figure of merit for determining the amount

of deuterium release from beryllium Increasing

Temperature (K) 400

Present data PISCES

Present data PISCES

Causey and Walsh TPE

0.01 0.1 1

0.01 0.1 1

Figure 8 Comparison of D/Be levels in beryllium codeposits with the O/Be levels in the same codeposits.

Reproduced with permission from Baldwin, M J.; Schmid, K.; Doerner, R P.; Wiltner, A.; Seraydarian, R.; Linsmeier, Ch.

J Nucl Mater 2005, 337–339, 590–594.

the time spent at lower baking temperatures did

not increase the amount of deuterium released from

the beryllium codeposits These results, along with

the retention level predictions, should make it

possi-ble to design baking systems for different areas of a

confinement device to control the accumulation rate

of tritium to a desired level

4.19.3.3 Mixed-Material Effects

A recent review of mixed-material effects in ITER62

provides background information on mixed-material

formation mechanisms and plasma–surface

interac-tion effects Here, the focus is on

beryllium-containing mixed-material surfaces (i.e., Be/C and

Be/W) and the conditions when one might expect

these surfaces to dominate the observed plasma–

surface interactions In addition to plasma

interac-tions with mixed-material surfaces, which will be

discussed here, other aspects such as changes to

ther-mal conductivity, material strength, and ductility, the

impact of impurities on material joints, etc., must also

be carefully evaluated

4.19.3.3.1 Be–C phenomena

Beryllium and carbon have been observed to begin

thermally interdiffusing at a temperature of around

500C,56 resulting in the formation of a beryllium

carbide layer However, beryllium carbide has also

been observed to form during energetic carbon ion

bombardment of beryllium surfaces at room

temper-ature.110 As mentioned in Section 4.19.3.1.2, the

change in the binding energy of the carbide molecule

affects the sputtering yield of both the beryllium and

carbon atoms In addition, the formation of beryllium

carbide also has a dramatic effect on the chemical

erosion properties of a carbon surface bombarded

with energetic beryllium ions.67,68,111

The presence of beryllium carbide on the surface of

a carbon sample exposed to deuterium plasma has been

shown to correlate with the reduction of chemical

erosion of the carbon surface.70 The speculation for

the cause of this effect is that the carbide enhances the

recombination of deuterium in the surface, thereby

lessening the amount of deuterium available to interact

with carbon atoms on the surface This is similar to the

impact of small amounts of boron carbide in a graphite

surface affecting chemical erosion.112 However, the

difference here is that instead of obtaining the carbide

through an expensive production technique, the

car-bide forms naturally as beryllium ions in the plasma

interact with the carbon surface

A systematic study of the time necessary to press chemical erosion of a graphite surface due tothe interaction with beryllium-containing plasma hasbeen carried out.69 Increasing the surface tempera-ture of the graphite was seen to have the biggestimpact on reducing the suppression time Increasingthe beryllium content of the plasma also reduced thesuppression time in a nonlinear fashion An increase

sup-of the incident particle energy was observed toincrease the time necessary to suppress the chemicalerosion of the surface, presumably due to an increase

in the removal of the carbide-containing surfacelayer A subsequent study showed that applyingheat pulses to a graphite surface interacting withberyllium-containing plasma, to simulate surfaceheating due to intermittent events, acted to reducethe time necessary for the carbide surface to form andsuppress the chemical erosion of the surface.1134.19.3.3.2 Be–W alloying

The existence of tungsten beryllide alloys (i.e., Be2W,

Be12W, and Be22W) is an excellent example of theimportance of mixed-material surface formation inplasma-facing components.114 Figure 9 shows thetungsten–beryllium phase diagram Each of the ber-yllides shown in the figure exhibits a lower meltingtemperature than one would expect from a tungstenplasma-facing surface If plasma containing berylliumimpurities interacts with a tungsten surface, there is

a possibility of these lower melting temperatureberyllide alloys being formed

In thermodynamic equilibrium, various beryllidealloys of tungsten have been observed to form,115andtheir reaction rates have been measured,116at tempera-tures in excess of 800C However, as was seen withberyllium carbide forming during plasma bombard-ment at lower temperature than expected thermody-namically, the concern exists that tungsten beryllidecould form at temperatures below 800C as well.Well controlled laboratory measurements in vac-uum117and in plasma simulators118have shown thatalthough thin, nanometer scale, Be2W layers form

at the interface between beryllium and tungstensurfaces, their growth below 800C is negligible Inaddition, above 800C, rapid beryllium sublimationfrom surfaces can act to limit the amount of beryl-lium available for reacting with tungsten and therebyalso limit the growth rate of the alloys In the presentlow wall temperature confinement devices, modelingshows that the divertor strike point locations are theonly areas where significant beryllide growth might

be expected and in these regions there does not

appear to be enough beryllium deposition to raise

significant concerns.119One caveat to these

predic-tions would be the existence of intermittent events

that raise the temperature of surfaces where

signifi-cant beryllium deposits are located, thereby possibly

allowing the optimized beryllide growth conditions

Another concern with regard to thin Be2W surface

layers on plasma-exposed tungsten is the impact of

these layers on tritium retention While a thin

Be2W surface layer is not likely to retain much

tri-tium itself, the thin beryllide surface layer could alter

the recombination characteristics of the bulk material

and change the accumulation rate of tritium within

the device To date, there is little or no data available

to address this issue

While it appears likely that the most serious issues

of tungsten beryllide formation may be avoided in

present confinement devices, the issues associated

with these alloys highlight the uncertainties and

impor-tance of understanding and predicting mixed-material

formation in plasma environments Mixed materialsoften interact with plasma in much different waysthan their elemental components In the case of theberyllium–carbon system (Section 4.19.3.3.1), themixed material appears to offer the potential for bene-ficial effects, whereas in the case of the beryllium–tungsten system, the mixed material appears likely to

be detrimental to the operation of the device Eachmixed-material system must, therefore, be individuallyevaluated to determine its potential impact on allaspects of operating surfaces in contact with plasma

4.19.4 Main Physical and Mechanical Properties

4.19.4.1 General Considerations

A comprehensive, although not recent, review ofthe science and technology of beryllium can befound in Beryllium Science and Technology.120

500 1000 1500 2000 2500 3000

3500 0

Weight percent tungsten

Figure 9 Phase diagram for the Be–W system Reproduced with permission from Doerner, R P.; Baldwin, M J.; Causey, R A J Nucl Mater 2005, 342, 63–67.

Several reviews have been published recently related

to use of beryllium in tokamaks and the status of the

investigations of the Be properties for the fusion

application.3,121–126 Various production and

proces-sing methods of beryllium metal fabrication have

been reviewed in Dombrowski.127 The majority of

methods are based on powder metallurgy and include

powder preparation from cast product by grinding

(i.e., attrition milling, impact grinding, ball mill

grinding); further powder consolidation (i.e., by cold

pressing (CP), cold isostatic pressing (CIP), vacuum

hot pressing (VHP), hot isostatic pressing (HIP)); and

possible additional mechanical treatment (e.g.,

extru-sion, rolling, forging) Beryllium protective armor

can also be produced by plasma spray (see Section

4.19.4.3) and vapor deposition

Several proposals were made at the beginning of

the ITER Research Programme during the ITER

Engineering Design Phase to develop a fusion grade

beryllium with high ductility, high resistance to heat

flux, and high radiation resistance However, it was

recognized that this development would require

sig-nificant efforts and could not be supported only by

requests from the fusion community

There are various beryllium grades, which have

been developed for different applications These grades

differ by chemical composition (BeO content,

impuri-ties), by method of powder preparation, by method of

consolidation, etc The nonexhaustive list of various

beryllium grades from the US and the Russian

Feder-ation is presented in ITER Materials Properties

Handbook (MPH).128Grades with similar

composi-tion are under produccomposi-tion in Kazakhstan and in China

We briefly discuss below some of the most relevant

physical and mechanical properties of beryllium,

in relation to its application as armor for PFCs

4.19.4.1.1 Physical properties

The physical properties of beryllium are summarized

in Table 2, which is taken from ITER MPH.128

These properties have been used for design and

performance assessments In addition to its low

atomic number, beryllium has several excellent

ther-mal properties that make it well-suited for heat

removal components The thermal conductivity is

comparable with that of graphite or CFC at low and

high temperatures but, in contrast to C-based

mate-rials, is not significantly degraded as a result of

neutron-irradiation The specific heat of beryllium

exceeds that of C-based materials typically by a

factor of 2 over the temperature range of interest

for operation However, Be has poor refractory

properties, such as low melting temperature andhigh vapor pressure The high heat capacity andgood thermal conductivity of Be can be used tomaintain low surface temperatures in PFCs duringnormal operation, but its low melting temperatureand high vapor pressure cause great design difficul-ties from the standpoint of survivability from off-normal events such as vertical displacement event(VDE), ELMs, disruptions, and runaway electronimpact (seeSection 4.19.6.2)

For the beryllium hexagonal close packed crystalstructure, the main physical properties, such as thecoefficient of thermal expansion, elastic modulus etc.have some anisotropy However, for the polycrystal-line grades these properties could be, in the firstapproximation, considered as isotropic Some anisot-ropy is also typical for the highly deformed grades.The physical properties (thermal conductivity, spe-cific heat, elastic modulus, etc.) in first approximationare the same for beryllium grades with similar BeOand other impurity content and they are produced bythe same fabrication method

4.19.4.1.2 Mechanical propertiesBeryllium is known to be a brittle material, with atypical elongation to failure in room temperaturetensile tests of roughly 0.8–6% For material withstrong anisotropy (e.g., rolled plate or sheet), elonga-tion in the rolling direction could be higher, but

in the transverse direction the elongation is

Table 2 Physical properties of beryllium

82 (800 C)

Specific heat (J kg1C1) 1900 Latent heat of fusion (kJ kg1) 1300 Latent heat of vaporization (kJ kg1) 3.66 10 4

Electrical resistivity ( mO cm) 4.4 (RT) Thermal expansion coefficient

RT, room temperature.

* Depending on quality of surface

typically significantly lower than 1% Recently, the

mechanical properties of beryllium have been

summarized in ITER MPH128and ITER Materials

Assessment Report (MAR).129

The mechanical properties of beryllium depend

on the production method used and they are sensitive

to a variety of factors including BeO and impurity

content (which varies from less than 1% to 2–3%

for various grades), method of powder preparation

(impact grinding, attrition grinding), method of

con-solidation, and further treatments The main problem

in using beryllium is its low ductility related to the

hexagonal-close-packed structure There is limited

slip in directions not parallel to the basal planes,

resulting in very small ductility perpendicular to

the basal direction Depending on the production

method, ductility of beryllium can be severely

anisotropic The grain size is an important factor in

determining the ductility of various beryllium

com-ponents Much of the fine grain size present in the

starting powder is retained during hot pressing at

1060C Without an oxide network, grain growth

occurs at a much lower temperature, about 800C

Among various beryllium grades, it was found that

grade S-65C VHP (production of Brush Wellman,

US) has the highest guaranteed fracture elongation

at room temperature (minimum 3%; typical is more

than 4–5%) This grade is produced using impact

grinding powder and has a guaranteed BeO content

<1% The level of impurities is also controlled

The high ductility of the grade is one of the

advan-tages of this material Because of the VHP production

method, there is some anisotropy of properties in

relation to hot pressing direction, but the differences

are not significant

As typical for all metals, the tensile properties of

beryllium depend on the testing temperature As the

testing temperature increases, a decrease of the

ulti-mate tensile and yield strength are observed

How-ever, rupture elongation increases with increasing

test temperature and could reach a value higher by

40–50% for temperatures around300–350C (see

as example data for grade S-65C VHP in the ITER

MPH128) A further increase in the test temperature

leads to a decrease of the elongation At temperatures

above 600C, the ductility depends on the impurity

content, mainly aluminum, which tends to segregate

at grain boundaries, impairing the mechanical

prop-erties By heat treatment in the temperature range

650–800C, aluminum can be combined with other

elements, mainly iron and beryllium itself, to form a

stable beryllide as AlFeBe However, the stable

beryllide dissolves progressively when heated at peratures>850C This last feature is important for

tem-the selection of the joining technology formanufacturing of the PFCs

Further details on mechanical properties, such

as creep and fracture toughness, can be found where (see, e.g., ITER MAR129)

else-4.19.4.2 Selection of Beryllium Grades forITER Applications

For ITER PFC applications, various commerciallyavailable beryllium grades from the United States(Brush Wellman Inc.) and from the Russian Federa-tion, listed in Table 3, were evaluated more than adecade ago as potential candidates during the ITEREngineering Design Activity (EDA)

The selection of the optimum grade for ITERPFC applications is driven mainly by the require-ments of ITER operation for structural integrity andstability against various thermal loads, and in partic-ular, the absence or minimization of macrodamage

It is believed that ion-induced and thermal erosion

at elevated temperatures is very similar for variousgrades of Be However, performance under high heatfluxes, especially under transient thermal loads such

as disruptions, VDE, and ELMs resulted in differentbehavior and damage mechanisms It is consideredthat the ease of joining beryllium to copper alloys(see Section 4.19.5) is not so sensitive to BeO con-tent, impurity levels, and method of consolidation,which are the parameters defining the grade of beryl-lium material

It should be noted that for tokamak applications(seeSection 4.19.6) beryllium is used in the form oftiles Some surface cracking of the tiles could beacceptable, if there is no macrodamage or delamina-tion along the surface of tiles, which leads to the loss

TGP-56 TShGT, DIP-30, TShG-200 VHP, vacuum hot pressing; HIP, hot isostatic pressing;

CIP, cold isostatic pressing.

because cracking could lead not only to enhanced

armor erosion, delamination, and loss of particles, but

also potentially to crack propagation to the heat sink

structure Neutron irradiation resistance is another

factor to be taken into account because it may affect

the thermal performance and structural integrity

Because of some of the uncertainties in the ITER

thermal loads, especially during transient events,

preference is given to beryllium grade(s) with

poten-tially higher resistance to transient thermal loads

The selection of the reference grades was made on

the basis of comprehensive assessment of the results

of various tests carried out during the ITER EDA

The detailed analysis is presented in ITER MAR.129

Among the various studies, the following shall be

mentioned:

 Screening low cycle fatigue test of 21 different

beryllium grades was performed in the past.130

It was shown that the grades with the best thermal

fatigue resistance are S-65C VHP, DShG-200,

TShG-56, and TShGT.Figure 10shows the results

of the comparative low cyclic thermal fatigue study

of different grades of beryllium

 Various grades of beryllium were also tested in

conditions simulating the disruption heat loads.131

The tests show that crack formation and behavior

after surface layer melting in different grades are

quite different For Be S-65C, all cracks stopped in

the molten zone, whereas for some grades the crackspropagated to the bulk of the sample

 Results of VDE simulation tests have beenreported in Linke et al.132,133Severe melting of Bewas observed for energy densities of 60 MJ m2(1 s pulse duration); however, no cracks wereobserved between molten and unmolten materialand in the bulk of unmolten parts for S-65CVHP grade

On the basis of the available data, Be S-65C VHP(Brush Wellman, US) was selected as the referencematerial on the basis of excellent thermal fatigue andthermal shock behavior, and for the good availabledatabase on materials properties, including neutronirradiation effects DShG-200 (produced in the Rus-sian Federation) was proposed as a backup, but thisgrade is no longer commercially available

Recently, China and the Russian Federation, thatare two of the seven International Parties engaged inthe construction of ITER, have proposed the fabrica-tion of additional first-wall grades as part of theirITER contribution The Russian Federation proposes

to use beryllium grade TGP-56-FW This grade isproduced by VHP in almost the final form of the tilesforeseen for the first wall The recent results ondevelopment of this grade have been reported inKupriyanov et al.134China proposes instead to use agrade called CN-G01135 that is produced from

0 0 500

1000

Side crack propagation depth (mm)

S-200F-H

S-200F (T) S-65C (T)

S-65C (L) DShG-200 (T)

Grades with best fatigue performance

S-65-H

TGP-56 S-200F (L)

94% S-65

98% S-65

Extruded (L) Extruded (T)

TShGT(T) TShG-56 (T) 1500

2000 2500

impact ground powder (similar to powder used for

S-65C grade) by VHP The grade is produced by

Ningxia Orient Non-Ferrous Metals Co Ltd

In order to accept these newly proposed beryllium

grades a specific qualification program is underway

4.19.4.3 Considerations on

Plasma-Sprayed Beryllium

In the past, plasma spraying was considered as a

high deposition rate coating method, which could

offer the potential for in situ repair of eroded or

damaged Be surfaces Development work was

launched during the early phase of the ITER

R&D Program in the mid-1990s.136 In the plasma

spray process, a powder of the material to be

depos-ited is fed into a small arc-driven plasma jet, and the

resulting molten droplets are sprayed onto the target

surface Upon impact, the droplets flow out and

quickly solidify to form the coating With recent

process improvements, high quality beryllium

coat-ings ranging up to more than 1 cm in thickness have

been successfully produced Beryllium deposition

rates up to 450 g h1have been demonstrated with

98% of the theoretical density in the as-deposited

material Several papers on the subject have been

published.136–138 A summary of the main

achieve-ments can be found inTable 4

However, based on the results available, the initial

idea of using plasma-sprayed beryllium for in situ

(in tokamak) repair was abandoned for several

rea-sons First was the complexity of the process and

requirements to control a large number of

para-meters, which affect the quality of the plasma sprayed

coatings Some of the most important parametersinclude plasma spray parameters such as (1) power,gas composition, gas flow-rate, nozzle geometry, feed,and spray distance; (2) characteristics of the feedstockmaterials, namely, particle size distribution, morphol-ogy, and flow characteristics; (3) deposit formationdynamics, that is, wetting and spreading behavior,cooling and solidification rates, heat transfer coeffi-cient, and degree of undercooling; (4) substrateconditions, where parameters such as roughness,temperature and thermal conductivity, and cleanli-ness play a strong role; (5) microstructure andproperties of the deposit, namely, splat characteris-tics, grain morphology and texture, porosity, phasedistribution, adhesion/cohesion, and physical andmechanical properties; and (6) process control, that

is, particle velocity, gas velocity, particle and gastemperatures, and particle trajectories Second,plasma-sprayed beryllium needs (1) inert gas pres-sure, (2) reclamation of the oversprayed powder(more than 10%), and (3) strict control of the sub-strate temperature The higher the temperature thehigher the quality of the plasma-sprayed coating, butunfortunately, an easy and reliable method to heatthe first wall to allow in situ deposition was not found.Finally, tools to reliably measure the quality of thecoating and its thickness are not available today and

a strict control of the coating parameters is difficult

to achieve

Thus, it was concluded that plasma-sprayed lium for in situ repair is too speculative for ITERwithout further significant developments Neverthe-less, this method still remains attractive and could beused for refurbishment of damaged components in

beryl-Table 4 Main achievements of ITER-relevant plasma-sprayed technology (summary of best results, not always achieved together)

Thermal conductivity (W mK1) Up to 160 at RT Depends on temperature of substrate, maximum achieved at

T  600–800 C with addition of H

Substrate temperature (C) >450 Very important for good strength, adhesion, and thermal

conductivity Keep in mind that CuCrZr temperature should not be higher than 500C for several hours due to overageing of CuCrZr Substrate preparation Negative

hot cell, albeit it may be cheaper to replace a

dam-aged component with a new one

4.19.4.4 Neutron-Irradiation Effects

Several authors have reviewed the properties of

neutron-irradiated beryllium for fusion applications

in the past.139–141Neutron irradiation leads to

com-plex changes in the microstructure, such as the

radiation-induced change of volume in beryllium,

which is dominated by the nucleation and growth of

He bubbles

There are two important pathways for gas

produc-tion One is the (n, 2n) reaction in which the9Be is

reduced to8Be, which then splits into two4He atoms

The second is the (n,a) reaction where the 9

Beabsorbs a neutron and then splits to form a4He and

a 6He The 6He rapidly undergoes a b decay to

become 6Li The 6Li then reacts with a thermal

neutron to produce 4He and 3H These processes

have been incorporated into the inventory code

FISPACT,142 which is used (see, e.g., Forty et al.143)

to estimate the generation rates of gas and other

reaction products in a tokamak

Helium generation has significant effects on the

properties of materials, especially at elevated

tempera-tures Helium is initially trapped within the beryllium

lattice in submicroscopic clusters At higher neutron

fluence massive helium-bubble-induced swelling

occurs, especially at elevated irradiation or postanneal

temperatures Because of the atomistic nature of the

helium bubble nucleation and growth, porous

beryl-lium microstructures, such as from powder metallurgy

or plasma spray technology, were not found to be

effective in releasing significant amounts of helium

under fusion reactor conditions.2

The maximum neutron-induced damage and

helium production expected in Be for ITER

first-wall applications (fluence of 0.5 MWam2) are

1.4–1.7 dpa and 1500 appm, respectively and the

expected irradiation temperatures are in the range

of 200–600C The maximum temperature is on

the surface of beryllium tile and depends on

thick-ness and heat flux Tritium production in beryllium

is expected to be about 16 appm Recently, Barabash

et al.144have analyzed the specific effects of

neutron-induced material property changes on ITER PFCs

foreseen during ITER operation

Typically, property changes induced by neutron

irradiation are investigated by exposing samples/

mock-ups in fission reactors However, the

differ-ences between the fission and fusion neutron spectra

are important to interpret and predict the effects.The key difference is transmutation production,which needs to be considered for the correct predic-tion of the material performance.145During irradia-tion in fission reactors, for example, the typical value

of the ratio (appm He per dpa) is 100–250, whereasfor a fusion neutron spectrum this value is 1000.Depending on operational temperature, the dpa or

He transmutation must be used as a reference tron damage parameter For beryllium, during low-temperature irradiation (<300C) the dpa value

neu-must be considered For high-temperature irradiation(more than 500C), the He generation must be

taken as the reference parameter

A detailed discussion on this subject is beyond thescope of this review We summarize only some of themain findings with emphasis on results for ITERrelevant grades Considerations of the effects of neu-tron irradiation of duplex Be/Cu alloy mock-ups areprovided inSection 4.19.5

4.19.4.4.1 Thermal conductivityFor S-65C Be grade irradiated up to 1025n m2(0.74 dpa) at 300C, the thermal conductivity

was found to be similar, within experimental error,

to that of the unirradiated material.146 Similarly,

no effect was seen for Be S-65C after irradiation

at 350 and 700C to a damage dose 0.35 dpa.147Significant changes in the thermal conductivitywere observed only for conditions that lead to sig-nificant changes of the beryllium structure, such asthe formation of a high density of radiation defects(especially at low irradiation temperature and highdose) or high (more than tens of percent) swelling.144Other physical properties (elastic modulus, coef-ficient of thermal expansion, etc.) are not influenced

by neutron irradiation (at least at the fluenceand temperature ranges relevant for the berylliumarmor for the ITER PFCs)

4.19.4.4.2 Swelling

It is well known that beryllium swells when irradiated

by neutrons, especially during high temperatureirradiation Reviews of the available swelling datafor different Be grades can be found elsewhere (see,e.g., ITER MAR,129Billone,139and Barabash et al.141).The computer code ANFIBE (ANalysis of FusionIrradiated BEryllium), has been developed andapplied in the past as an interpretative and predictivetool148for the prediction of beryllium swelling Thedriving force for the swelling is the presence of

He, which forms He bubbles, especially during