Page 420 - Fundamentals of Magnetic Thermonuclear Reactor Design

398 Fundamentals of Magnetic Thermonuclear Reactor Design

13.5.3.1 Single-Stage Annealing
One batch of specimens was irradiated in the SM-2 reactor at 160°C up to 0.2–
0.6 dpa [12]. The specimens were then subjected to a 10-h annealing at 350°C
and 400°C. The other batch was irradiated in the same reactor at 150°C up to
2.5 dpa and annealed for 10 h in vacuum at 300°C and 350°C. Reference speci-
mens, irradiated specimens and irradiated-and-annealed specimens were tensile
tested.
The obtained experimental results suggest that annealing is an effective pro-
cess, allowing irradiated specimens to recover their plasticity properties almost
completely.
This conclusion is valid for the whole range of doses up to ∼2 dpa. For the
Cu–Cr–Zr IG alloy, the effect of annealing is not great, due probably to the high
mobility of the Cr atoms, stabilising the clusters of radiation-induced defects.
The annealing effect on the thermal and electrical conductivity, determined
by the concentration of transmutation products, was insignificant, as transmuta-
tion processes are beyond the annealing control.

13.5.3.2 Irradiation–Annealing–Irradiation Cycle
It has been proved experimentally that irradiation–annealing–irradiation cycles
do not lead to any additive embrittlement of either pure copper or the GlidCopAl
25 IG alloy. The effects of a radiation-induced hardening and embrittlement of
an alloy due to a 0.0014 dpa irradiation followed by a 24-h annealing at 350°C
then followed by a 0.013 dpa irradiation are comparable to the effect of an
one-off up to 0.0154 dpa irradiation of the same alloy. It seems that a recovery
annealing of a copper alloy helps develop a microstructure that is relatively
resilient to further irradiation [13]. Then, perhaps, a periodic annealing should
be regarded as an effective way to improve the durability of copper heat-con-
ductive structures and in-chamber components of a fusion reactor.

13.6 MATERIALS FOR THE VACUUM CHAMBER AND SUPPORTS

The prime constructional material for the ITER vacuum chamber, divertor
and blanket is the 316L(N) IG–grade austenitic stainless steel. It differs from
the basic 316L(N) grade, historically employed in nuclear power engineer-
ing, in that it has a lower content of impurities, such as cobalt, niobium and
tantalum (because of the induced radioactivity minimisation requirement).
The steel is characterised by high plasticity and fracture toughness, resis-
tivity to cyclic fatigue, and the ferritic phase low content. In addition, it is
highly fabricable and retains its mechanical properties if subjected to weld-
ing or soldering [14].
In ITER operational conditions (an irradiation dose of up to 20 dpa), the
316L(N) IG steel shows an attractive combination of radiation characteristics.
Its plasticity does not fall below 10% up to a dose of ∼10 dpa [15]. No critical
swelling or helium embrittlement occurs even with an intense transmutation