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Making Tungsten Work
R.E. Nygren, R. Raffray, D. Whyte, M.A. Urickson, M. Baldwin, L.L. Snead
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10.1016/j.jnucmat.2010.12.289
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Journal of Nuclear Materials
Please cite this article as: R.E. Nygren, R. Raffray, D. Whyte, M.A. Urickson, M. Baldwin, L.L. Snead, Making
Tungsten Work, Journal of Nuclear Materials (2010), doi: 10.1016/j.jnucmat.2010.12.289
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Making Tungsten Work - ICFRM-14 Session T26 Paper 501 Nygren et al.
Making Tungsten Work
R.E. Nygren*1, R. Raffray2, D.Whyte3, M.A Urickson1, M. Baldwin4, L.L. Snead5
1
2
Sandia National Laboratories, Albuquerque, NM, USA
International Thermonuclear Experimental Reactor Organization, Cadarache,
FRANCE
3
Plasma Science and Fusion Center at MIT, Cambridge, MA, USA
4
University of California, San Diego, La Jolla, CA, USA
5
Oak Ridge National Laboratory, Oak Ridge, TN, USA
*Corresponding Author:
Richard. E. Nygren
renygre@sandia.gov
Sandia National Laboratories
Albuquerque, New Mexico, USA
1-505-845-3135, fax 1-505-845-3130
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Making Tungsten Work - ICFRM-14 Session T26 Paper 501 Nygren et al.
Making Tungsten Work
R.E. Nygren *1, R. Raffray2, D.Whyte3, M.A Urickson1, M. Baldwin4, L.L. Snead5
1
2
Sandia National Laboratories, Albuquerque, NM, USA
International Thermonuclear Experimental Reactor Organization, Cadarache,
FRANCE
3
Plasma Science and Fusion Center at MIT, Cambridge, MA, USA
4
University of California, San Diego, La Jolla, CA, USA
5
Oak Ridge National Laboratory, Oak Ridge, TN, USA
Abstract: Tungsten is the plasma facing material of choice in several design studies for
DEMOs and in development programs for advanced plasma facing components. Use of
tungsten in ITER for the divertor and consideration of a full first wall of tungsten have
increased the pace of research in fusion on tungsten. This paper characterizes the critical
issues in making tungsten work as a plasma facing material for a DEMO and cites past
work as well as current experiments, modeling and materials and component
development.
Keywords: fusion, divertor, tungsten, high heat flux
*
Sandia is a multiprogram laboratory operated by Sandia Corporation, a Lockheed Martin Company, for the United
States Department of Energy’s National Nuclear Security Administration under contract DE-AC04-94AL85000
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1
Introduction
We are gaining a better understanding of the path forward to Plasma Facing Components
(PFCs) for a fusion DEMO through design studies, interdisciplinary working groups, and
preparation for ITER (International Thermonuclear Experimental Reactor). Tungsten
(W) is the plasma-facing material of choice in several design studies for DEMOs and
development programs for advanced PFCs [1-2]. "Making tungsten work" implies
successful development of all the needed aspects of power handling, plasma-materials
interactions, etc. to produce successful PFCs with tungsten as the plasma facing material.
Design studies provide insights on issues, define a set of self consistent requirements and
typically project aggressive solutions for PFC technology. The detailed requirements for
PFCs depend on the fusion power, magnetic configuration, and scheme for power
handling. We also draw from two recent informative US activities. The first is a
workshop with international, interdisciplinary participation that focused on the
development of divertors for a tokamak DEMO summarized in a recent paper [3]. Other
recent papers [4-6] review development and application of tungsten and tungsten alloys
for fusion. The second US activity, called ReNeW or the Research Needs Workshop,
engaged US researchers from all disciplines in the US fusion program in defining the
scientific and engineering basis for a (US) development path in fusion for the next 15-20
years [7].
2
Uncooled and Water-cooled W PFCs
Recent results on damage to W from plasma surface interactions and increasing
understanding of the potential threats to ITER’s PFCs from plasma disruptions and ELMs
(edge localized modes) are giving us a better picture of likely operating conditions for
PFCs in a DEMO. Reevaluation of the associated heat loads [8] led to a redesign of the
ITER first wall. 1 The new information will likely have impacts for a DEMO. Research
for ITER has focused effort on the development, fabrication and testing of W parts,
1
The previous design, in which the shape of the “conforming” first wall followed a magnetic flux surface
without added protection from poloidal or toroidal limiters, was based on the assumption that the heat load
was primarily line radiation from the plasma or charge exchange neutrals but without pathways to the first
wall for charged particles.
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including fabrication and testing of solid W tiles [9] and W coatings on carbon tiles [10]
for the ITER-like divertor experiment [11] in JET (Joint European Torus). Earlier
supporting R&D for ITER [12] explored several types of W armor joined to water-cooled
copper alloy (CuCrZr) heat sinks that included EU mockups with (a) W “macro-brush”
armor made by cutting larger tiles into unit cells of 5x5 mm or (b) 5-mm-thick plasma
sprayed W armor, EU and Japanese mockups with CVD (chemical vapor deposited) W
tiles, US development of targets with W rods joined to CuCrZr heat sinks, and further
work in Japan on embedded W rods. The early US work initially identified the need to
subdivide or “castellate” W tiles to reduce thermal stresses.
Other fusion experiments (ASDEX, Max Planck Institute of Plasma Physics; Alcator CMOD, Massachusetts Institute of Technology, and TEXTOR, Forschungszentrum Jülich)
have already used W PFCs without active cooling. ASDEX progressively increased its
wall coverage with W to 100% of the PFCs in 2007 [13], but also has used boronization
to mitigate contamination of the plasma core with W and offered advice on use of W in
future fusion applications [14]. At C-MOD, researchers have installed a row of divertor
tiles made of 4-mm-thick W plates, slotted at the back and standing on edge and bound
together by a longitudinal bolt through the center of the plates, with the objective that
these tiles survive 5 second pulses with absorbed heat loads up to 15 MW/m2 without
melting or damage [15]. Also, TEXTOR has operated with solid and coated W limiters
[16].
For a heat sink deployed in the HT-7 tokamak, Luo and collaborators prepared a copper
(Cu) heat sink that had a vacuum-plasma-sprayed coating with a graded composition
from a Cu-rich zone to pure W [17]. Applications with graded composition of Cu to W
have also been developed in the US [18]. Tokunaga and colleagues have worked on W
coating of graphite and carbon fiber composites for about a decade [19]. There is also
interest in W coatings on the steels proposed for the first wall of a DEMO [20].
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3
High Temperature W PFCs
Higher operating temperatures, robustness and lifetime, and integration with a tritium
breeding blanket are the primary differences between ITER PFCs and those appropriate
for a DEMO. A desire for high efficiency energy conversion with a Brayton cycle in a
DEMO and high expected heat loads in the divertor motivates the preference for a
helium-cooled refractory structure. A copper is not suitable for these high temperatures
(and high radiation doses) and ferritic steels lack the required thermal conductivity (~100
W/m-K) combined with adequate strength at temperature (>1000°C). Some other
concepts for divertors and first walls use liquids metals. Europe has a strong regional
effort (including Russia, Japan, US, Sweden and Finland) with work on tungsten for
ITER plus development of helium-cooled W divertor mockups and DEMO studies. High
heat flux testing of helium-cooled mockups with W armor or W structure has been done
in electron and ion beam facilities in the US, Europe, Russia and Japan. Linke’s
excellent recent overviews describe the development, characterization and testing of
plasma facing materials and components for future fusion devices [6,21].
He-cooled PFCs require a high mass flow rate for adequate heat removal. This means
high density at acceptable flow velocities and therefore high pressure. Several design
studies specify helium-cooled PFCs with a He pressure in the range of 8-10 MPa and a
refractory structure or refractory armor. A primary constraint is the thermal stress
generated by the asymmetric (one-sided) heating that makes designs with long tubes and
coolant paths unworkable. The applications described below illustrate some design
solutions, but at this point the designs are fairly complex assemblies and the applications
for DEMO would require thousands of such parts.
3.1 HEMJ Divertor Module
Norajitra and co-workers [22-23] developed a He-cooled modular divertor with jet
cooling (HEMJ in Fig. 1) in which the plasma facing armor is a flat W tile brazed to a Walloy thimble and recently tested a mockup with an array of 9 modules. Helium at 10
MPa arriving at 600˚C and exiting at 700˚C cools the thimble, which is coupled to piping
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Making Tungsten Work - ICFRM-14 Session T26 Paper 501 Nygren et al.
of oxide-dispersion-strengthened (ODS) Eurofer (steel) through a W-to-steel transition
piece.
In earlier work, the US company, Thermacore, Inc., developed a modular concept with a
a refractory alloy tube brazed to hexagonal W headpiece. Its hemi-spherical interior well
held a packed bed of W particles, also brazed, through which helium flowed outward
from the center. This refractory module removed a heat load of 6 MW/m2 over an area of
2160 mm2. In more recent tests, a dual channel He-cooled tungsten heat sink by
Thermacore, Inc., for which the possibility of channel-to-channel flow instabilities was of
interest, reached 34.6 MW/m2 [24].
3.2 T-Tube
The T-tube configuration is a natural consequence of placing the inlet and outlet close
together at the center of a heat sink. For the refractory T-tube in Figure 2, from the
ARIES-CS design study [25], He in a coaxial W-alloy cartridge flows outward through
an array of holes that create a dense series of He jets to cool a W-alloy tube with a brazed
saddle block of castellated W armor. The transition piece (noted in Figure 2), which has
a graded composition, can be joined at one end to W and at the other to steel. The design
parameters are: ~6 kg/s He per m2 of divertor surface, pressure drop of ~0.11 MPa
through the jets (greatest flow resistance), surface heat flux of 10 MW/m2, inlet and outlet
temperatures of 570 and 700°C respectively, maximum W alloy temperature of 1240°C
(below 1300°C limit assumed for recrystallization) and maximum combined primary and
secondary stresses of ~342 MPa. Initial R&D on this concept at the Georgia Institute of
Technology confirmed an effective heat transfer coefficient of ~0.04 MW/m2-K [26].
Refractory foams may also offer unique possibilities for fusion PFCs. One example is a
CVD processes used by Ultramet, Inc. to form a structural shell of tungsten and then coat
an interior network of carbonized foam ligaments with tungsten. The coated ligaments
form an open foam with a large fraction of open volume, as opposed to many packed
beds used for porous media. Also, the process produces an integral joint, rather than a
braze, between the shell and the porous foam. A tungsten tube, 15 mm in diameter with a
38-mm-long internal mesh of tungsten-coated ligaments, sustained a peak heat flux of
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Making Tungsten Work - ICFRM-14 Session T26 Paper 501 Nygren et al.
~22 MW/m2 with He at 4 MPa flowing at 27g/s, inlet and outlet temperatures of 40˚C
and 91˚C respectively, and a pressure drop of ~0.07 MPa. However, this pure tungsten
tube failed by brittle fracture during the cooling phase of the test [27].
Porous media can enhance heat transfer or act as a host for other materials that cover their
surface, such as liquid metals or boron. One idea for a coating for a first wall is a porous
tungsten mesh with boron impregnated and covering the surface [28]. A tungsten mesh
was initially proposed as the host for lithium in the Liquid Lithium Divertor now installed
in the National Spherical Torus Experiment [29].
4
W Properties and Development
Pure tungsten (or perhaps an alloy with other constituents only at grain boundaries) is
typically preferred as a refractory plasma facing armor for PFCs because of its low
sputtering yield. The other structure of a PFC might use a tungsten alloy, or some form
of W yet to be developed primarily to improve the ductility, joined to another material
specified for manifolds. This is done due to the weight, cost and limitations of W-based
parts, and the requirement to couple the power removal systems in the first wall and
blanket. Ferritic steels are the leading class of materials for these related applications;
however there is some concern regarding their combination of relatively low thermal
conductivity and high desired service temperatures in a DEMO.
W is a low-activation material with a high melting point, high thermal conductivity, and
low thermal expansion. Its recrystallization temperature and ductile-brittle transition
temperature (DBTT) bound the range of useful operation temperature. Both are of
critical concern in using these materials for PFCs. For example, recrystallized material,
even in the surface layer of PFC armor, would likely increase cracking during transient
heat loads, propagation of cracks during cooling, and the related threats of (a) the
formation and loss of loose particulates from the surface and (b) crack growth in
structural components.
4.1 Heat Loads
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Excellent thermo-mechanical performance is a requirement for refractory PFCs in
DEMO, and our design-related testing typically includes thermal cycling tests. An
example of desired criteria for structural materials in a divertor include creep strength
(~55 MPa for minimum time to rupture of 20,000 hours at 1200°C), thermal conductivity
(~100 W/m-K at 1200°C), and DBTT (~300 °C, unirradiated) [3].
An open and important question is the severity of the transient heat loads (and electromagnetic loads for plasma disruptions) that a DEMO must accommodate. These are
poorly defined at present, but again, experience for ITER in defining ELMs and plasma
disruptions shows the nature of such loads.
In major plasma disruptions, the plasma looses stability, dumps energy rapidly
(milliseconds), shifts position and, as the current flowing in the plasma quenches, image
currents generated in the surrounding structure (and flowing across the high magnetic
fields) can cause brief but severe mechanical loads in the mounting hardware of the first
wall and divertor. The thermal energy density carried by the plasma edge outside the
separatrix 2, parallel to the field and near the outer and inner divertor targets respectively
is 100-600 MJ/m2 and 130-780 MJ/m2, and the projected loads on the divertor plates are
4-25 MJ/m2 and 7-40 MJ/m2 with rise and decay times of 1.5-3 ms and 1.5-6 ms. In
more rapid Vertical Displacement Events (VDEs), ~20-30 MJ/m2 lands on the upper wall
modules in ~0.1 ms or downward onto the divertor dome, baffle or lower wall modules in
~0.3 ms.
During ELMs, the plasma’s overall position remains stable, but particles carry energy in
short bursts outward from the plasma edge. Controlling the ELMs in ITER is a hot topic
of research. For controlled ELMs, the maximum energy densities on the divertor targets
are 0.3 MJ/m2 and 0.5 MJ/m2 for the outer and inner plates respectively with a deposition
time of 0.25-0.5 ms and frequency of 20-40 Hz. Uncontrolled ELMs, with heat loads of
10 MJ/m2 and 6 MJ/m2 to the outer and inner plates at a frequency of 1-2 Hz, present
more of problem.
2
The separatrix is a surface that separates the core plasma from the edge plasma. The magnetic field lines
(iso-surfaces) in the space of the core plasma close on themselves and do not intersect the PFCs, whereas
the opposite is true outside the separatrix.
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We can also look at what heat loads might be acceptable. Using data from high heat flux
tests in various facilities that varied from very short power pulses to sustained thermal
cycling tests, Linke et al. correlated these data with a thermal diffusion parameter (heat
flux multiplied by square root of time) to identify thresholds below which boiling,
melting, cracking or roughening did not occur [6].
Although there are rules regarding the use of materials with low ductility, use of a brittle
material in PFCs where long lifetimes and robustness are desirable seems quite imprudent.
Consequently the development and clever combination of W-based materials with
adequate ductility for fusion applications is a very important goal for fusion technology.
4.2 Development and Testing
Industrial activity in refractory metals is strong world-wide, but the development efforts
directed toward materials development in applications for fusion are limited. European
efforts lead (contributions are noted below) and there is research in Japan [30,31]. US
activities are only intermittent, primarily through government grants to small businesses.
FZK (Forschungszentrum Karlsruhe) and PLANSEE screened currently available Wbased materials in a study of the following five W rod materials from PLANSEE: pure
W; WL10, which has 1% lanthanum oxide, in two different conditions; W doped with
0.005% potassium, here called WVM; and WL10 with 1% Re; plus plates of pure W,
WL10, and WVM [32]. Mo-Ti-Zr (TZM) rod and plate were used as references. Using
impact bending tests (in vacuum), the study showed the influence of microstructural
characteristics like grain size, anisotropy, and texture, or the influence of chemical
composition. Their data show encouraging early observation regarding creep strength
and thermal conductivity for unirradiated WL10 that satisfy the design goals noted
previously. Regarding recrystallization, the recrystallization temperature should be
defined with a minimum time equal to material’s operation life, which is 20,000 hours for
the design noted previously. Results for pure W showed a recrystallization temperature
of 1300 °C; in other studies at 1300 °C, WL10 had not recrystallized after ~2000 hours.
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With respect to ductile behavior, the effect of the stress state is well known and notched
specimens have higher DBTTs. For the W rod materials, all show brittle fractures below
600 °C. The fractures of those with notches made with a cutting saw were fully ductile at
800 °C and showed delamination-to-ductile fracture at 750 °C. Fractures of those with
notches made by electro-discharge-machining were fully ductile at 900 °C and showed
delamination-to-ductile fracture of 850 °C. The WVM rod specimens were fully ductile
at 1000 °C (delamination-to-ductile transition at 950 °C). WL10, WL10opt,
W1Re1La2O3 rod materials do not show fully ductile fracture up to 1000°C. All plate
materials (including TZM) exhibited severe delamination with the brittle-to-delamination
transitions at 150 °C and 450 °C for TZM and W respectively. Delamination (a
characteristic failure with separation among layers or filaments) occurs in the anisotropic
materials or textured materials, but not in the recrystallized refractory materials with
equiaxed grains.
Earlier studies on commercial weld electrode materials showed DBTTs above 900 °C,
and all the W rod materials in the study noted above exhibit brittle fracture below about
600 °C, which by comparison is an improvement although still probably inadequate for
DEMO applications. Plate materials (rolled) in general perform worse than (rolled) rod
materials due to the different microstructures that develop during material production.
The type of machining, through the stresses generated with the tool contact, also affects
the resulting properties of fabricated parts. For example, in studies done for the HEMJ,
electrical discharge machining of W tiles produced more micro-cracking of the surface
than finish grinding [33]. Dry milling, i.e. grinding with a CBN grinding wheel, took 30
minutes compared with 2 hours milling with a coolant, and the surface was not
contaminated for the subsequent brazing process. The effects of specific chemicals in
electromachining have also been reported [34]. Optimum fracture behavior can possibly
reached only by avoiding machining and by aligning the grains along the contour of the
according part. For example, the result of the fabrication studies for the thimble in the
HEMJ is a recommendation for a combination of deep drawing and twisting. In an
intriguing new application, researchers at the Karlsruhe Institute of Technology
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developed tungsten feed stock and used powder injection molding to form a block and
disk as a prelude to future production of HEJM thimble armor blocks [35].
Improvement of material processing, suppression of recrystallization, and a slight
strengthening in creep resistance are benefits of lanthanum-oxide particles in W.
However, the already inadequate fracture behavior of pure tungsten deteriorates further
with the addition of lanthanum oxide (and to a lesser extent with potassium). These
obviously promote delamination, probably by weakening the grain boundary cohesion.
Kurishita has shown improvements in critical mechanical properties of W with ultrafinegrain size and nano-dispersoids of TiC made using a thermo-mechanical treatment [31].
At this time this lab-scale operation produces only small lots of material.
4.3 Radiation Effects
The conclusions to date are from studies of unirradiated material. Neutron irradiation, as
well as implantation of energetic particles (D,T,He) from the plasma and related damage
to the surface layers of PFCs will affect the material properties. Effects on embrittlement
and the requirements that might limit the operating temperatures of PFCs are of particular
concern, since neutron irradiation tends to increase the DBTT and also produce
transmutations [36-38]. There are planned irradiations on W samples and even some
planning for near term irradiations of W-armored targets that will be water-cooled in
post-irradiation high heat flux tests [39] to provide data for ITER.
An important aspect of the quest to mitigate the brittleness of tungsten-based materials is
to differentiate between neutron damage that is intrinsic and stable versus effects that can
be manipulated by alloying. For example, stacking fault tetrahedra can form in copper
(an FCC material) within the damage cascade and are then stable with respect to
annealing. There is also concern that some type of intrinsic damage can form in BCC
materials such as W and molybdenum (Mo), which has many similarities to W. In Mo,
-4
small cavities were clearly detected even for a damage dose below 10 dpa. With
increasing dose, hardly any change in average cavity size was seen, in contrast to the
results for Fe where cavity sizes increased with increasing dose. These observations
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suggest that the in-cascade vacancy clustering may be significant in neutron-irradiated
Mo, in agreement with suggestions from Molecular Dynamics simulations [40].
5
Plasma Materials Interactions
Concerns in plasma materials interactions (PMI) for a DEMO include erosion, retention
of tritium, creation of dust and the effects of ion and neutron damage, and particularly the
effects of He that will be implanted in the surfaces of PFCs. Excellent general reviews
by Philipps [41] and by Roth et al. [42] provide more information.
Tungsten is the worst impurity in a plasma due to power losses by line radiation from
many possible charge states and the potential for its accumulation in the core plasma.
Yet we prefer W as refractory armor for PFCs based on its high sputtering yield and the
premise that we will manage the conditions of the edge plasma in a DEMO to limit the
feedback of W from the PFCs to the core plasma. Modeling of erosion generally includes
a distribution of the energies of ions (D,T, He, impurities) appropriate for the plasma
edge and a contribution from energetic neutrals, as in a recent analysis for an all-metal
ITER [43].
We are also seeing plasma interactions with W that we do not understand, and the
projection to a high-power DEMO of the behavior of the plasma edge also carries
significant uncertainty. At this point, we assume our DEMO will have repeatable
plasmas (without the range of operational scenarios for research devices), and that we can
manage these plasmas with some acceptable transient events, that are not yet well
characterized. Thus the assumed thermal loads to the PFCs will vary in time, as will their
surface temperatures and thermal gradients, and their surfaces will change over time as
they accumulate damage from both ions and neutrons.
The evolving nature of these surfaces in a DEMO complicates our ability to understand
and model these PFCs. The group of PMI issues discussed separately below are actually
interdependent and several also have a strong dependence on temperature.
5.1 Tritium
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Tritium will be closely monitored and its retention in PFCs is a concern. One type of
concern for safety (and licensure) are postulated accidents that could result in a breach of
the vacuum vessel and release of vapor volatized when ingress of the outside atmosphere
contacts hot PFCs with retained tritium. Pathways that remove tritium from the fuel
cycle, especially those that are hard to monitor, affect both the safety and economics of a
DEMO. There are recent experiments for ITER simulating tritium in W and even the
effects of neutrons [44]. Some effects of tritium, i.e., transmutation into He and electrical
charging of particulates, differ from the effects of hydrogen and deuterium (D) in metals.
But most of our knowledge about tritium retention comes from studies of the retention of
D in metals which is generally low, especially at elevated temperatures, but enhanced by
neutron and ion damage. References [42] and [43] discuss this, and Reference [45]
shows data from laboratory experiments and observations in ASDEX-Upgrade. In
considering the retention of tritium in W, most data are for fluences in the range of 1025
m-2, with a few points extending to 1x1026 where the retention is still increasing, and
there is a single point at 1 x 1027. Almost all these data were collected at relatively low
temperature (<500ºC). DEMO will have both higher temperatures and much higher
fluences.
5.2 Dust, Blisters and Fuzz
Dust produced from flaking films or dislodged particulates has a high ratio of surface
area to volume (concern for tritium retention and explosions); the volume of dust may be
difficult to monitor, and the sources are not well understood [46]. One possible
mechanism for forming W dust was observed directly by researchers at Nagoya
University. They found ejected grains of W dislodged from a sample after accumulation
of He from doses in their linear plasma source NAGDIS-II had caused extensive He
bubbles to form at and weaken the grain boundaries of tungsten grains adjacent to the
free surface [47]. Another concern is that particles falling into the plasma may cause an
accumulation of W or trigger a disruption.
Other possible mechanisms for formation of dust include arc tracks, motion of melted
material, condensation of vaporized material, and breaking of blisters formed from
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implanted He and D. At low temperature (200-300ºC) bombarding W with pure
deuterium (D) at~40-60 eV tends to make surface blisters and increase the retained D. In
raising the temperatures for such exposures, the blisters tend to disappear and the
retention level decreases. Adding a few percent He to the D at 200-300ºC greatly reduces
the blisters and the retained D. However, at ~800-1000ºC, the surface develops a nanostructured (fuzz) morphology. The tungsten “fuzz” (Figure 3) was a startling discovery
by researchers at Nagoya using NAGDIS-II and at the University of California, San
Diego using their linear plasma source PISCES, and later found also in a W sample
exposed to D and He plasmas in LHD [47-49]. He stabilizes the growth of voids and
promotes growth of the tendril structure. Subsequent studies in PISCES show that the
growth of this nano-morphology is persistent and occurs in W at the temperature ranges
of interest for ITER and DEMO. Although a structure of this type would be unlikely to
survive on a PFC with a high heat load, some active driving force that will modify the
surface clearly exists, and the issue remains: What modification will take place?
All of the phenomena noted above as well as the effects of neutron irradiation are likely
to play a role in the evolution of the surfaces of W PFCs. Two important areas of PMI
where currently planned research in ITER will not address research needs for DEMO are
PFCs operating with (1) at high neutron damage and (2) high temperature. Also, the
combination of these conditions will likely exacerbate issues associated with an evolving
microstructure and material properties. Lab scale experiments, modeling and additional
experiments in fusion devices will be critically important in extending our understanding
of plasma-materials interactions with W. The integrated testing in a fusion device with
hot walls, and therefore a fundamentally different regime in physical chemistry due to the
temperature dependences of the processes, is completely unexplored territory.
6
Making Tungsten Work (Conclusion)
Developing tungsten for fusion means developing robust actively-cooled tungsten-based
PFCs. This implies a parallel and coordinated program of lab scale experiments,
component development, experiments in fusion devices and modeling. Moreover, the
experiments and modeling should go forward together such that test conditions and
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results are measured with the accuracy and completeness useful for benchmarking
models with which we can confidently predict the performance of DEMO PFCs. Below
are the basic elements of such a program.
• development of W-based materials, including production, fabrication, joining and
neutron damage
• investigation of PMI issues
• measurements of the plasma edge
• development and deployment of PFCs (confirmation of design, goals for
performance, QA, high heat flux testing, etc.)
• experiments with large areas of W PFCs and operation with hot walls
• modeling of materials and the evolution and effects of damage
• modeling of the plasma edge
• benchmarking of edge and materials models
• modeling of component performance and integrated testing
• benchmarking of predictive performance models that integrate plasma edge,
materials evolution and component performance
As noted in the Introduction, the ReNeW activity (Research Needs Workshop)[50] was
an extensive effort within the US to define the gaps in the current program with respect to
future development. The basic elements noted above were recognized in this activity and
discussions continue within the US community of fusion researchers to develop
recommendations for the detailed research and development needed in these areas.
Individual experiments should contribute to a useful set of complementary data worldwide. Consider, for example, choice of materials. Tungsten is available in various
wrought and powder metallurgy forms. Alloys as well as ultra-fine grain W are being
considered as alternatives to pure (brittle) W. Additives that form grain boundary
precipitates, improve mechanical properties and retard recrystallization are considered
beneficial. But minor constituents can also affect the formation of damage sites and
trapping of helium and hydrogen and alter plasma–materials interactions. We see
exciting possibilities, such as the ‘nanodispersoids’, to improve materials, but we will
14
Making Tungsten Work - ICFRM-14 Session T26 Paper 501 Nygren et al.
also face pressure to pare down the possibilities and focus on a relatively few choices to
follow through in more intensive stages of development.
The question is how we move forward to make tungsten work and when. Important
contributing factors will be recognition of the need to move forward with fusion nuclear
science and technology, adequate commitment and funding, appropriate leadership and
good international cooperation and collaboration.
Let us accept here that a likely stage on the development path toward DEMO is some
complement of confinement facilities in parallel with ITER that have strong mission for
development of technology and one or more meaningful experiments with hot walls. We
are already seeing proposals for “satellite” tokamaks, upgrades and D/T devices for
developing fusion nuclear science and technology.
As more powerful fusion experiments require actively-cooled PFCs and missions demand
better efficiency and more running time each year, the requirements for PFCs will
become more stringent. Hot actively-cooled W-based PFCs are a significant step in
technology that includes both the development of the relevant materials and further
development and confirmation of the technology for heat removal. Since any new piece
of hardware is also a threat, the intent to introduce new PFCs in some device implies the
need to mitigate adverse impacts through adequate design and extensive preparation and
qualification. This is certainly necessary so that W-based PFCs can be successfully
deployed at this intermediate stage in the path of fusion development. * We will likely
have stepwise progress, as with the ITER-like divertor experiment, and perhaps the
development of a specific facility, before we are likely to have all the conditions to
produce relevant data for DEMO.
Technology rather than physics will pace the development of fusion. As world attention
focuses more on fusion as an energy source, and decision-makers must champion this
development, we will need to show supporting science and engineering that confirm that
the technology for the path forward is credible and achievable. Confirming that we can
make tungsten work is an important strategic goal.
*
This theme of integrated development is explained further in DEMO Divertor Development, by Nygren et
al., 9th Int. Symp. on Fusion Nuclear Technology, Dalian, China, October 2009, to be published in FED.
15
Making Tungsten Work - ICFRM-14 Session T26 Paper 501 Nygren et al.
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21
Making Tungsten Work - ICFRM-14 Session T26 Paper 501 Nygren et al.
Figures
1. He-cooled W divertor module and 9 module array. A recent adaptation (in box, lower
right) covers a larger surface area and individual thimbles are used only in the area of
peak heat load Figure courtesy of Forschungszentrum Karlsruhe. One column
2. T-tube in the ARIES-CS divertor. X-section shows He path. He flows up through
central hole in stalk into inner tube in cartridge, the up through holes to form He jets
at top of cartridge. Return flow is along outside annulus of cartridge and into the two
outlet channels in the stalk. Courtesy of Center for Energy Research, UCSD One
column
3. Formation of tungsten “fuzz” during simultaneous bombard of He and D at 1200 °C
in PISCES, courtesy of the Center for Energy Research, University of California, San
Diego One column
22
Fig. 1
Flat W
armor
brazed
to
W alloy
thimble
He flow
600˚C
10 MPa
W-to-steel
transition
(brazed)
Outlet
700˚C
ODS EUROFER
structure
Nygren
*Figure 1 is saved as p1 in file: [501]_Nygren_Fig1_2_3.JPEG
Fig. 2
Nygren
*Figure 2 is saved as p2 in file: [501]_Nygren_Fig1_2_3.JPEG
Fig. 3
Nygren
*Figure 3 is saved in as p3 in file: [501]_Nygren_Fig1_2_3.JPEG