The simulation of disruptions (<20 MJ m~2, t = 5 ms) on VPS-W and W alloys irradiated to a dose of 0.2 dpa at 350 and 750 °C resulted in heavy melting of the material241 but yielded no measurable degra­dation by neutron irradiation. This is understandable because the decrease in thermal conductivity, which is the most important material parameter for melting experiments, is almost negligible for the given irradia­tion conditions.217 Because of the continuous modifi­cation of the material composition by transmutation described above, with increasing levels of Re and Os, the thermal conductivity and the related melting threshold power density is expected to decrease steadily.

In further investigations applying ELM-like loads on pure W and W-La2O3 irradiated to 0.6 dpa at 200 °C, the crack pattern generated after irradiation exhibited an increased crack density in combination with a smaller crack width. Furthermore, in W-0.2% Re SC that was exposed to the same neutron irradia­tion conditions and exhibited no crack formation in the nonirradiated state, cracks were formed along the crystallographic planes (see Figure 9).177 The effect was more obvious in the results for the SC material, but in both cases the observed degradation was a result of mechanical property changes and a rise of the DBTT in particular. This would indicate a rise in the threshold temperature for crack forma­tion (see also Section 4.17.4.3.3), which has not been verified yet. For the evaluation of the material’s performance in DEMO, both higher transmutation rates and significantly higher temperatures that are expected to stimulate defect recovery have to be taken into account.

4.17.4.3.1 Thermal fatigue on irradiated W components

Information on the thermal fatigue resistance of W components is limited to the experience obtained in two irradiation campaigns for ITER which reached neutron doses of 0.15 and 0.6 dpa at 200 °C. Refer­ence and irradiated actively cooled mock-ups with W—1% La2O3 as the PFM were exposed to 1000 cyclic steady state heat loads at power densities up to 18 MW m~2 213,216,217

The results obtained indicate that at these neutron fluences the material changes occurring in tungsten do not have any significant influence on the component’s performance. However, mock-ups based on the macro­brush design experience a degradation ofthe maximum achievable power density from 18 to 10MWm~2, which is related to neutron embrittlement and subsequent cracking failure of the Cu-heat sink mate­rial. In contrast, monoblock mock-ups show identical high-level performance before and after irradiation, which makes it the favored design for ITER.

Despite these positive results, based on the irradiation-induced mechanical property changes outlined above, the use of tungsten in any highly stressed component at low temperatures <500 °C

has to be avoided.108