Liang Gao

Fusion power is transitioning from a primarily physics-driven challenge to one dominated by materials and component qualification. Many sub-components have performed acceptably in relevant environments, yet deployment timelines keep slipping because qualification is treated as a prerequisite for construction rather than a process that matures alongside operation and is validated through it. This paper argues for a structuredpragmatism approach, in which a first-of-a-kind (FOAK) fusion power plant designed from the outset as a materials and components substantiation platform. Drawing on system-level assumptions for a European gigawatt-class stellarator power plant such as GIGA, the paper shows that credible,licensable baselines already exist. The remaining uncertainties are not primarily about material feasibility but about quantified performance in the integrated system. They hinge on practical, system-level constraints, manufacturing route, joining technique, quality assurance, local radiation environment, lifetime requirements, tritium control, inspection strategy, replacement schedule, and waste route, that determine whether a material choice is deployable in a physical reactor. The paper proposes a qualification framework for early fusion plants operating under bounded uncertainty, combining bounded operating envelopes, staged commissioning, embedded materials surveillance, neutronics-informed evidence ranking, manufacturing traceability, and risk-informed licensing into a single materials-intelligence substantiation loop. Its central conclusion is that fusion materials qualification cannot be completed before the first power plant exists; it must be completed through that plant’s controlled, evidence-generating operation

Fuel retention in plasma-facing components (PFCs) is a critical issue in future nuclear fusion reactors operating with Deuterium-Tritium (DT) regarding nuclear safety and fulfillment of the T cycle. However, during DT plasma operation, highly energetic neutrons will induce damage in the lattice of W PFCs causing enhanced fuel retention in defects or traps. Laser-Induced Breakdown Spectroscopy (LIBS) is a potential tool to monitor the T-content in situ in PFCs of future nuclear fusion devices. This article presents an ex situ study on pre-damaged W material after D plasma exposure to qualify the method and mimic conditions expected in a reactor. ITER grade W samples were displacement-damaged by 10.8 MeV W ions to a damage dose of 0.23 dpa and exposed to low temperature deuterium plasma at low energy in PlaQ. The resulting deuterium concentration was analyzed by using 3He Nuclear Reaction Analysis (depth resolution of ≈150 nm) as a well-established method, and LIBS (picosecond laser pulses, depth resolution of 15 nm). The sample with the highest deuterium concentration showed a deuterium-rich zone up to a depth of 1.13 μm using both techniques. This is close to the expected W ion-induced damage depth of ≈1 μm. The results imply that LIBS as an in situ technique for tritium monitoring could be a viable option for a reactor.

This comment examines a recent study claiming the observation of deuterium (D) supersaturated surface layer (SSL) in tungsten samples exposed to D plasma in the PISCES-A plasma device at ion energies as low as 45 eV/D (Nishijima et al 2023 Nucl. Fusion 63 126003). Applying SDTrimSP simulations and recalling the previous study on SSL formation, herein we want to emphasize that the proposed model in the mentioned paper relies on a number of strong assumptions and that many of the observations made can be more easily rationalized by the presence of impurities. The present comment will be conducive to avoid some possible misunderstanding on the SSL formation mechanism.

Tungsten fiber-reinforced tungsten (W_f/W) has been developed to improve the fracture toughness of W materials, as demonstrated in previous studies (2019 Nucl. Fusion 59 086034; 2021 Mater. Sci. Eng. A 817 141361). In the present study, we focus on the performance of the developed W_f/W materials under fusion-relevant test conditions and further demonstrate their use as plasma facing materials in future fusion reactors. Specifically, one set of W_f/W samples was exposed to Ne plasma to investigate the erosion resistance against plasma sputtering, in comparison to the reference ITER-grade W sample. In addition, deuterium (D) retention in the plasma-exposed W_f/W samples was studied via thermal desorption spectroscopy. Furthermore, laser thermal shock tests were performed on W_f/W to simulate the transient heat load condition and to investigate the material performance under extreme heat flux. With increasing porosity, W_f/W exhibits lower mass loss (net erosion) after Ne plasma exposure. Though porous, W_f/W composites unexpectedly show a comparable D retention to the reference bulk W, which is attributed to the openness of the pores in the matrix. Thermal shock testing results indicate a similar cracking threshold (0.38 GW m⁻², 1 ms) as compared with that of ITER-grade W materials. However, due to the lower thermal conductivity of porous matrix W_f/W, under extremely high loading conditions (1.6 GW m⁻², 2 ms) surface melting was observed. The present work demonstrates the great potential of the porous matrix W_f/W for future fusion application.

The influence of the transmutation element rhenium (Re) on defect evolution in tungsten (W) during irradiations with heavy ions and deuterium (D) plasma was investigated. Rolled W and W–5Re alloy (Re concentration 5 wt.%) were irradiated with 500 keV iron ions to 0.06 and 0.6 dpa (displacement per atom), and deuterium plasma at 38 eV/D to a fluence of 2.2 × 10²⁵ D m⁻² at ∼373 K. The results of Doppler broadening spectroscopy by the positron annihilation technique indicate that more or larger vacancy-type defects were produced in W than in the W–5Re during Fe ion irradiation, suggesting the important role of Re on inhibiting the migration and/or clustering of vacancies. The much smaller and shallower blistering in W–5Re than in W after the same D plasma exposure points to the pinning effects on dislocations by Re. The total retained D amounts in W–5Re and W materials are comparable, but with very different concentration profile at corresponding depth. This is explained by the blister formation exerting influence on the D inward diffusion. Demonstrating the important role of Re on the defect evolution in W under irradiation, the present work provides an enhanced understanding on the possible effects of neutron irradiation on the performance of plasma-facing materials in future fusion devices.