Conference Agenda

The EUROfusion Consortium is an association of Laboratories across 28 European countries in charge of implementing the European Roadmap to Fusion Electricity. Development and qualification of neutron tolerant materials is one of the missions in the Roadmap, with neutron irradiation and subsequent post irradiation examination representing a key challenge requiring substantial time and financial resources, and long-term vision for the codification of the results into reactor design codes.

In this paper, we provide an overview of the EUROfusion contribution to the investigation of the neutron irradiation effects in different commercial and advanced tungsten grades provided by different partners within EUROfusion as well as via international collaboration programmes, including divertor relevant joints. The presentation summarizes the main results and lessons learned over the 2014-2026 period, and it is sub-divided into five sections covering: (i) commercial ITER-specification tungsten; (ii) advanced tungsten grades achieved by alloying or innovative production methods; (iii) impact of thermal-neutron shielding on the irradiation damage; (iv) performance under very high irradiation temperature; (v) irradiation of W-CuCrZr and W-Cu-CuCrZr joints. The effects of neutron irradiation are assessed in terms of the modification of mechanical properties, and supported by microstructural investigations to explain and interpret the obtained results.

The divertor is the key component for power exhaust in modern magnetic confinement fusion devices. Historically, the engineering efforts were focused on high-heat-flux (HHF) technologies to cope with the severe thermal loads and surface erosion caused by intensive particle bombardment. Even today, HHF technology remains one of the most critical challenges. Consequently, the technology for HHF target has been often regarded as the synonym of divertor technology per se.
The situation began to change with the ITER divertor. Neutron irradiation effects and the thermal impact of slow transients had to be taken into account. Moreover, design of the steel components such as cassette body or target supports required a proper structural verification against codes as a pressure vessel under nuclear environment. These technical peculiarities drove dedicated technology R&D efforts for the steel pressure components as well as HHF targets. Still, the envisaged nuclear loads and cooling conditions of the ITER divertor are modest allowing a considerable safety margin.
When it comes to the divertor of a power plant scale reactor, the design and technology options are subject to various restrictive high level requirements and technical constraints derived from essentially different loading and boundary conditions, which pose a profound impact on design and technology options for the entire divertor system. In the framework of EUROfusion, the ‘Divertor’ project (WPDIV) has been addressing these issues for over a decade with its comprehensive design and R&D activities taking the European demonstration reactor (EU-DEMO) as reference.
Recently, a substantially revised divertor design was created on the occasion of the new EU-DEMO baseline. The underlying design features were motivated by the high level requirements to ensure technical compatibility with DEMO scale reactors and enhance viability towards power plant applications. Technology gaps were identified and prioritized. Extensive R&D items were derived from the design and implemented into the WPDIV program where dozens of European laboratories and companies are involved.
In this contribution, starting with the unique challenges and outstanding design issues raised specifically for the divertor of a tokamak-type fusion power plant, we present potential power exhaust scenarios, the current status of design study based on the European approach, the latest R&D progress for the entire divertor system, remote maintenance scheme and estimated pumping performance. Finally, the implications for future R&D endeavors are discussed.

Accurate characterisation of neutron spectra in irradiation facilities is essential for reliable activation studies, detector calibration, and interpretation of irradiation experiments. In water activation experiments, particular attention must be paid to the high-energy component of the neutron spectrum, as threshold reactions significantly influence the production of short-lived radionuclides.

This work presents a computational and methodological study aimed at characterising the neutron field in the Inner Irradiation Snail of the KATANA water activation facility, with a focus on preparation for experimental validation using activation dosimetry foils. The study is based on detailed Monte Carlo simulations performed with the MCNP code, providing estimation of the neutron flux distribution and energy spectrum within the irradiation position.

The calculated neutron spectrum reveals a mixed field with a measurable fast neutron component extending into the MeV range, which is particularly relevant for water activation reactions such as ¹⁶O(n,p)¹⁶N and ¹⁷O(n,p)¹⁷N. Based on the simulated spectrum, a systematic analysis of suitable activation reactions was performed to define an optimised set of dosimetry foils. Emphasis was placed on threshold reactions capable of probing the high-energy part of the spectrum, including ²⁷Al(n,α)²⁴Na, ⁵⁸Ni(n,p)⁵⁸Co, ⁵⁴Fe(n,p)⁵⁴Mn, among others.

Selection criteria for the dosimetry materials included reaction threshold energy, cross-section magnitude and shape, half-life of activation products, gamma emission characteristics, and suitability for high-resolution gamma spectrometry. The aim was to ensure adequate sensitivity across the relevant energy range while maintaining practical feasibility for irradiation and post-irradiation analysis.

In addition to the spectral and reaction analysis, practical aspects of experimental implementation were considered. These include foil geometry and positioning within the irradiation channel, expected activity levels, counting requirements, and handling procedures for activated materials. Special attention was given to radiological safety considerations associated with the experimental setup. The aluminium positioning rod used for sample placement becomes activated during irradiation, producing short-lived gamma-emitting radionuclides (e.g., ²⁴Na). To mitigate the resulting dose rates during sample retrieval, dedicated shielding and handling procedures were designed and implemented to ensure the safe extraction of the irradiated foils.

The study also outlines the planned methodology for the subsequent experimental campaign, including irradiation conditions and activity measurements using HPGe detectors. The identified set of activation reactions is expected to enable detailed assessment of the fast neutron component, which is essential for accurate modelling of water activation processes, establishing the KATANA water activation facility as a benchmark, and enabling integral cross-section measurements.

The neutron capture reaction ¹⁸⁶W (n,γ) ¹⁸⁷W is of high importance to fusion applications, due to its effect on neutron balance in the breeder assembly. Neutrons, especially at the lower energies in the expected first wall neutron spectra in ITER and DEMO, can be absorbed in tungsten. This induces activity, which is important for waste management and disposal as well as workplace safety. Furthermore, the decay results in accumulation of rhenium in the pure tungsten first wall components, eventually leading to creation of rhenium precipitates and embrittlement of the material. Cross section validation was performed in three integral experiments with different neutron spectra. First experiment used a thermal neutron beam built at HK-3 horizontal channel of the LVR-15 research reactor. The results were compared to calculations using IRDFF-II, JENDL-5 and JEFF 4 libraries. The difference between experiment and calculation is bellow experimental uncertainty for all libraries. Second experiment was performed in a vertical channel at the edge of the core of the LVR-15 research reactor with a thermal reactor neutron spectrum. In this case the sensitivity is the highest in the epithermal energy region. The neutron spectrum shape was calculated using an MCNP model of the reactor. The spectrum shape was combined with neutron flux measured using neutron activation monitors and activation cross sections from the IRDFF-II, JENDL-5 and JEFF3.3 libraries. The difference between experiment and calculation using IRDFF-II activation library is 24 %, the experimental uncertainty is 4 %. The high neutron flux in the experimental reactor allowed measurement of additional tungsten reactions, including ¹⁸⁷W (n,γ) ¹⁸⁸W. In high neutron flux a significant portion of the radioactive ¹⁸⁷W undergoes additional capture reaction ¹⁸⁷W (n,γ) ¹⁸⁸W. This leads to the eventual production of osmium instead of rhenium, influencing the impact of neutron irradiation on the mechanical properties of future tungsten components in fusion reactors. The difference between experiment and calculation for this reaction is 67 % for the JENDL 5 library. The last experiment was performed using a ²⁵²Cf(sf) standard neutron field. The experimental spectrum averaged cross sections (SACS) were compared to calculated SACS. Calculations were performed using activation cross sections from IRDFF-II, JENDL-5 and JEFF 4 libraries. In this case the difference between experimental SACS and calculated SACS in ²⁵²Cf PFNS is bellow experimental uncertainty for all libraries.

In future thermonuclear devices such as DEMO displacement damage by 14 MeV fusion neutrons will be created in the plasma-facing materials while they are exposed to high fluxes of ions and neutrals of hydrogen isotopes (HIs) at elevated temperatures. It was shown in several studies in the past twenty years that HI retention and transport will be dominated by trapping at the defects created by the neutron irradiation [1]. Moreover, the synergism between displacement damage creation and presence of HIs needs to be considered for realistic predicting of fuel retention and transport in future fusion devices.

In this study several W polycrystalline samples were irradiated by 10.8 MeV W ions at 800 K and 1000 K at different damage doses for D retention analysis, XRD and TEM analysis. Samples were after W irradiation exposed to 300 eV/D ions at 450 K and analysed by ³He nuclear reaction analysis. The purpose is to study the evolution of deuterium retention and defects at high temperatures and higher doses at low damage dose rates of the order of few times 10^-5 dpa/s using a continuous beam. We observe a saturation of D concentration above 0.2 dpa for samples irradiated at 800 K, which means that there is also saturation of open volume defects observed previously by Schwarz-Selinger at 300 K [1]. In the case of 1000 K we observe an increase in D concentration between the sample irradiated at 0.2 dpa and 0.8 dpa, which might indicate that we are not yet in D concentration saturation at 0.8 dpa accompanied with the occurrence of bigger voids compared to 800 K case. Detailed analysis of microstructure was performed by STEM/TEM on all samples. The dislocation density is the highest at 0.02 dpa and for higher damage doses the dislocation density decreases becoming less dense and dislocation lines become larger at higher damage doses of 0.4 dpa and 0.8 dpa. Void size analysis shows that we have voids of about 1.5 nm size in all samples irradiated between 0.2 dpa to 0.8 dpa at 800 K. Void size increases for samples irradiated between 0.2 dpa to 0.8 dpa at 1000 K to about 2 nm and shows voids also for sample irradiated at 0.02 dpa where at 800 K no voids were observed. Simultaneous W irradiation and D ion exposure was performed at 1000 K and damage dose of 0.8 dpa. D depth profile analysis after additional D ion exposure after the irradiation to probe the created damage confirmed the old results [3] where increased D retention was observed as compared to only W irradiation case. TEM analysis of these samples revealed increase in void size by 60 % which indicates that even very low D presence during W irradiation at 1000 K increases void size.

References:

[1] A. P. Persianova, & A. V. Golubeva, A Review. Phys. Metals Metallogr. 125, 278–306 (2024)

[2] T. Schwarz-Selinger, Mater. Res. Express 10, 102002 (2023)

[3] S. Markelj, et al., Nucl. Fusion 59, 086050 (2019)