10.1 Criteria for Selection of PFM
189
W as PFM, it seems possible to avoid the W accumulation in the plasma center with
appropriate plasma control. However, operational temperature window for W used
as PFM is limited because of quite low allowable impurity level and avoidance of
material damage caused by melting and embrittlement. Therefore, the tolerant power
load to W would be less than that for C. For C, in spite of its large erosion, damages
like cracking and fracture could be avoided by using carbon fiber enforced carbon
materials (CFC). Benefit of C is that erosion could be repaired by C deposition by
CVD and PCVD processes for example. Low erosion of W is advantage for materials
lifetime. However, heat shock damage and materials degradation, i.e. loss of ductility
or increased of DBTT by neutron irradiation and recrystallization of W due to cyclic
power load are concerned. Low neutron activation of C is another advantage making
maintenance and repairing easier.
Since the shape of plasma-facing surface (PFS) could not be ideally smooth both
in toroidal and poloidal directions, high heat load at the edges of plasma-facing
tiles cannot be avoided. Self-shaping, i.e. erosion at salient region and deposition
at concave region, makes PFS flatten and consequently erosion would be reduced.
High erosion and deposition of C-wall could be compensated by the self-shaping.
Still depositions at plasma shadowed and remote areas are concerns for C. In this
respect, W is better owing to less erosion.
Tritium retention is one of the critical issues. Low retention is better and has been
one of the most important criteria for the selection of PFM of ITER. However, as
described in the previous chapter, easy recovery would be more important for T fuel
self-sufficiency. For C-wall, most of T is retained in deposited layers and near-surface
regions of eroded area, while bulk retention is dominant for W-wall. In any PFM, T
recovery must be done, although required techniques are different.
In summary, W seems to be a very good PFM, if the central accumulation of high
Z impurities could be avoided by suitable transport control. “Prompt deposition”
of high Z atoms is very promising because it reduces the erosion and long-range
transport, consequently deposition at remote area is suppressed. Under the same
power load condition, the net power load is likely less for high Z materials than that
for low Z materials because of their higher reflectivity both of energy and particle,
which must be confirmed in the future.
10.2 Concerns on W Usage as PFM
W seems a good candidate PFM in aspects of low erosion (and hence low deposition)
and low T retention. However, various concerns in actual use of W as PFM of burning
plasma remain as described in this section. ITER could be one of the most important
test beds for W usage as PFM. In particular, it would be one of the most important
tasks to study the effects of W on the operation of burning plasma. Unfortunately,
however, concerns on materials aspects in long term usage could not be tested in
ITER and requires separate examination to realize W PFM in a reactor. Power load
tests using linear plasma machines would help this.
189
W as PFM, it seems possible to avoid the W accumulation in the plasma center with
appropriate plasma control. However, operational temperature window for W used
as PFM is limited because of quite low allowable impurity level and avoidance of
material damage caused by melting and embrittlement. Therefore, the tolerant power
load to W would be less than that for C. For C, in spite of its large erosion, damages
like cracking and fracture could be avoided by using carbon fiber enforced carbon
materials (CFC). Benefit of C is that erosion could be repaired by C deposition by
CVD and PCVD processes for example. Low erosion of W is advantage for materials
lifetime. However, heat shock damage and materials degradation, i.e. loss of ductility
or increased of DBTT by neutron irradiation and recrystallization of W due to cyclic
power load are concerned. Low neutron activation of C is another advantage making
maintenance and repairing easier.
Since the shape of plasma-facing surface (PFS) could not be ideally smooth both
in toroidal and poloidal directions, high heat load at the edges of plasma-facing
tiles cannot be avoided. Self-shaping, i.e. erosion at salient region and deposition
at concave region, makes PFS flatten and consequently erosion would be reduced.
High erosion and deposition of C-wall could be compensated by the self-shaping.
Still depositions at plasma shadowed and remote areas are concerns for C. In this
respect, W is better owing to less erosion.
Tritium retention is one of the critical issues. Low retention is better and has been
one of the most important criteria for the selection of PFM of ITER. However, as
described in the previous chapter, easy recovery would be more important for T fuel
self-sufficiency. For C-wall, most of T is retained in deposited layers and near-surface
regions of eroded area, while bulk retention is dominant for W-wall. In any PFM, T
recovery must be done, although required techniques are different.
In summary, W seems to be a very good PFM, if the central accumulation of high
Z impurities could be avoided by suitable transport control. “Prompt deposition”
of high Z atoms is very promising because it reduces the erosion and long-range
transport, consequently deposition at remote area is suppressed. Under the same
power load condition, the net power load is likely less for high Z materials than that
for low Z materials because of their higher reflectivity both of energy and particle,
which must be confirmed in the future.
10.2 Concerns on W Usage as PFM
W seems a good candidate PFM in aspects of low erosion (and hence low deposition)
and low T retention. However, various concerns in actual use of W as PFM of burning
plasma remain as described in this section. ITER could be one of the most important
test beds for W usage as PFM. In particular, it would be one of the most important
tasks to study the effects of W on the operation of burning plasma. Unfortunately,
however, concerns on materials aspects in long term usage could not be tested in
ITER and requires separate examination to realize W PFM in a reactor. Power load
tests using linear plasma machines would help this.