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The aim of the present study is to extend the knowledge about the formation and thermal stability of vacancy-type defects in tungsten under neutron irradiation, thereby mimicking the temperature and neutron flux expected in the ITER divertor. Neutron irradiation of single crystal tungsten, W(100), in the temperature range 600-1200 °C is performed up to 0.12 dpa. Positron annihilation spectroscopy is employed to detect the presence of open volume defects, while hardness tests are applied to relate the irradiation-induced defects with the modification of mechanical properties. Rationalization of the experimental results is enhanced by the application of a kinetic Monte Carlo simulation tool, applied to model the microstructural evolution under the neutron irradiation process. The relation between radiation microstructure and hardness is explained via a dispersed barrier model.

The M4F project brings together the fusion and fission materials communities working on the prediction of radiation damage production and evolution and their effects on the mechanical behaviour of irradiated ferritic/martensitic (F/M) steels. It is a multidisciplinary project in which several different experimental and computational materials science tools are integrated to understand and model the complex phenomena associated with the formation and evolution of irradiation induced defects and their effects on the macroscopic behaviour of the target materials. In particular the project focuses on two specific aspects: (1) To develop physical understanding and predictive models of the origin and consequences of localised deformation under irradiation in F/M steels; (2) To develop good practices and possibly advance towards the definition of protocols for the use of ion irradiation as a tool to evaluate radiation effects on materials. Nineteen modelling codes across different scales are being used and developed and an experimental validation programme based on the examination of materials irradiated with neutrons and ions is being carried out. The project enters now its 4th year and is close to delivering high-quality results. This paper overviews the work performed so far within the project, highlighting its impact for fission and fusion materials science. Peer Reviewed L. Malerba a,*, M.J. Caturla b, E. Gaganidze c, C. Kaden d, M.J. Konstantinovi´c e, P. Olsson f, C. Robertson g, D. Rodney h, A.M. Ruiz-Moreno i, M. Serrano a, J. Aktaa c, N. Anento j, S. Austin i, A. Bakaev e, J.P. Balbuena b, F. Bergner d, F. Boioli k, M. Boleininger l, G. Bonny e, N. Castin e, J.B. J. Chapman l, P. Chekhonin d, M. Clozel m, B. Devincre k, L. Dupuy g, G. Diego a, S.L. Dudarev l, C.-C. Fu g, R. Gatti k, L. G´el´ebart g, B. G´omez-Ferrer n, D. Gonçalves g, C. Guerrero a, P.M. Gueye n, P. H¨ahner i, S.P. Hannula o, Q. Hayat p, M. Hern´andez-Mayoral a, J. Jagielski m, N. Jennett p, F. Jim´enez a, G. Kapoor d, A. Kraych h, T. Khvan e,q, L. Kurpaska m, A. Kuronen r, N. Kvashin j, O. Libera s, P.-W. Ma l, T. Manninen o, M.-C. Marinica g, S. Merino a, E. Meslin g, F. Mompiou t, F. Mota a, H. Namburi s, C.J. Ortiz a, C. Pareige n, M. Prester u, R.R. Rajakrishnan t, M. Sauzay g, A. Serra j, I. Simonovski i, F. Soisson g, P. Sp¨atig v, D. Tanguy h, D. Terentyev e, M. Trebala o, M. Trochet g, A. Ulbricht d, M.Vallet g, K. Vogel d, T. Yalcinkaya w, J. Zhao r a Centro de Investigaciones Energ´eticas, Medioambientales y Tecnol´ogicas (CIEMAT), Madrid, Spain b Universidad de Alicante, San Vicente del Raspeig, Spain c Karlsruher Institut für Technologie (KIT), Karlsruhe, Germany d Helmholtz-Zentrum Dresden-Rossendorf Ev (HZDR), Rossendorf, Germany e Studiecentrum voor Kernenergie / Centre d’´Etude de l’´Energie Nucl´eaire (SCK CEN), Mol, Belgium f KTH Royal Institute of Technology, Stockholm, Sweden g Universit´e Paris-Saclay, Commissariat `a l’´Energie Atomique et aux ´Energies Alternatives (CEA), Gif-sur-Yvette, France h Institut Lumi`ere Mati`ere (ILM), Centre National de la Recherche Scientifique, Lyon, France i Joint Research Centre (JRC)- European Commission, Petten, the Netherlands j Universitat Polit`ecnica de Catalunya, Barcelona, Spain k Laboratoire d’Etude des Microstructures (LEM), Centre National de la Recherche Scientifique, Chˆatillon, France l United Kingdom Atomic Energy Authority (UKAEA), Culham, UK m Narodowe Centrum Badan Jadrowych (NCBJ), Swierk, Poland n Normandie Univ, UNIROUEN, INSA Rouen, CNRS, Groupe de Physique des Mat´eriaux, 76000 Rouen, France o Aalto University, Espoo, Finland p Coventry University, UK q Universit´e de Li`ege, Belgium r Helsingin Yliopisto, Helsinki, Finland s Centrum Vyzkumu ˇReˇz S.R.O. (CVR), ˇReˇz, Czech Republic t Centre pour l’´Elaboration Elaboration de Mat´eriaux et pour l’´Etude des Structures (CEMES), Centre National de la Recherche Scientifique, Toulouse, France u Institut za Fiziku, Zagreb, Croatia v Paul Scherrer Institut (PSI), Villingen, Switzerland w Middle East Technical University (METU), Ankara, Turkey

The thermal stability and the structure of solute-vacancy clusters formed by neutron irradiation are studied by means of positron annihilation spectroscopy and hardness measurements of post-irradiation annealed reactor pressure vessel steels with high and low Ni contents. Two distinct recovery stages were observed and assigned to (a) the dissolution of vacancy clusters at about 650 K, and (b) the dissolution of solute-vacancy clusters at about 750 K. In steels with high Ni content, hardening mainly recovers during the second stage. Atomistic and coarse grain models suggest that during this stage, the removal of vacancies from vacancy-solute clusters leads to complete cluster dissolution, which indicates that solute clusters are radiation induced.x.

T91 steel samples were irradiated in both a neutron and a neutron plus proton irradiation environments, and the resulting microstructure was studied using transmission electron microscopy. Three samples were irradiated with neutrons and protons to a cumulative dose of 0.88, 2.07, and 4.35 dpa at temperatures of 254 °C, 260 °C and 320–350 °C, respectively and two samples were irradiated with neutrons to 0.6 and 1 dpa, at a temperature of 290 °C for comparison. Radiation induced dislocation loops were observed in both groups of samples. In all samples irradiated up to 1 dpa, irrespective of irradiation conditions, dislocation loops were segregated near the line dislocations, forming a clear heterogeneous distribution. In the high dose sample, on the other hand, they were uniformly distributed over the grain interior and the middle dose sample showed a mixed character. Voids of similar average size were only observed in the middle and high dose samples, and their appearance seems to be correlated with the onset of homogeneous loop distribution. The observed defect properties are rationalized on the basis of modeling results.

A probabilistic fracture model is applied to irradiation assisted stress corrosion cracking effect, assuming that the oxidized part of stainless steel sample plays an essential role in the crack initiation, propagation and the sample failure. The Weibull statistical distribution of time-to-failures, estimated through the correlation with the statistical distribution of oxide strengths, including both the surface oxide layer and oxidized grain boundaries, fully describe the experimental scatter obtained in the constant load time-to-failure tests. Large failure uncertainties in these type of tests originate from an intrinsic stochastic behavior of the oxide cracking due to subcritical crack propagation process.

Vacancy-solute clustering in neutron irradiated Fe–Cr alloys with various concentrations of Cr and minor solutes (Ni, Si and P) were studied by using coincidence Doppler broadening spectroscopy and small angle neutron scattering techniques. The results from both experiments, supported by an object kinetic Monte Carlo model, show in a very consistent way the existence and formation of vacancy-CrNiSiP clusters that play detrimental role in irradiation hardening. Similar solute cluster number density of about 30–50 × 10 16 c m − 3 and an average diameter of about 1 nm were estimated for all alloys containing minor solutes, irrespectively of the chromium content. In Fe9Cr ferritic and Fe9Cr ferritic/martensitic alloys, with significantly reduced concentration of minor solute elements, the main defects are vacancy clusters, with an average cluster size of about 10 and 2 vacancies, respectively. Large concentration of α ’ -precipitates was observed in Fe14Cr(NiSiP). However, both vacancy clusters and α ’ -precipitates provide significantly less impact to hardening in comparison to vacancy-CrNiSiP clusters. The fact that vacancy clustering in Fe9Cr ferritic alloy resembles that of pure iron suggests that Cr solutes may play lesser role in irradiation hardening of ferritic alloys and steels than previously believed.

The mechanical properties of EPDM polymers, degraded as a result of extensive thermal and radiochemical aging treatment, are studied. The focus is given to dose rate effects in polymer insulation materials extracted from industrial cables in use in Belgian nuclear power plants. All studied mechanical characteristics such as the ultimate tensile stress, the Young's modulus, and the total elongation (or elongation at break) are found to be strongly affected by the irradiation dose. The ultimate tensile stress and Young's modulus are clearly exhibiting the dose rate effect, which originated from oxidation mediated interplay of polymer cross-linking and chain scission processes. The change of crossover between these two processes is found to be gradual, without critical dose rate or temperature values. On the contrary, the total elongation is observed not to be sensitive neither to irradiation temperature nor to the dose rate. Both cross-linking and chain scission seem to affect the total elongation in a similar way by reducing the average polymers chain length. This idea is confirmed by the model which shows that all total elongation data as a function of irradiation time can be reproduced by varying a single parameter, the pre-exponential factor of the irradiation rate constant.

The expected degradation of mechanical properties of structural materials under irradiation (i.e. “in operation”) in nuclear components represents a significant challenge for the design to account the impact of long-term irradiation effects. Therefore, a development of scientific and engineering expertise to understand and possibly control the severe impact of harsh neutron irradiation on materials is one of the tasks in the current material’s research agenda. As it is well known from long-standing experience with fission systems, radiation embrittlement is caused by nano-scale features that obstruct plasticity mediated by dislocations. The present contribution highlights recent research and development efforts addressed towards the assessment of the interrelation between nano-structural features and hardening induced by neutron irradiation in fusion structural high-Cr ferritic/martensitic steels. In this work, the impact of doping by three chemical elements (nickel, silicon and phosphorus) on the microstructural evolution under neutron irradiation at 300 °C and 450 °C is studied experimentally as well as by computer simulations. The evolution of microstructure is assessed by transmission electron microscopy and object kinetic Monte Carlo simulations. The latter method utilizes a new approach enabling to follow both nano-scale irradiation defects and micro-segregation zones causing the formation of solute rich clusters, detectable by atom probe tomography. The results of the present work clearly point out that Ni–Si segregation alters the spatial and size distribution of dislocations loops already at as low dose as 0.1 dpa. The physical mechanisms behind the impact of minor alloying elements are discussed.

The mechanical properties of spallation target components exposed to combined effects of proton and neutron irradiations and in contact with liquid metal provide important information for the assessment of structural component integrity, which is crucial for the design of accelerator driven reactor concepts such as the MYRRHA reactor. In this study we perform tensile tests on T91 steel samples extracted from the MEGAPIE, and from the TWIN-ASTIR experiment. The tests are performed at different temperatures as well as with and without the contact with liquid metal. In both groups of samples we observed significant influence of liquid metal on the tensile properties, in particular reduction of total elongation. The influence of different conditions in two irradiation programs on the mechanical properties, such as irradiation temperature fluctuations, the presence of neutron/proton irradiation, with and without the contact with lead-bismuth eutectic, different flux and fluence, are also discussed.

The mechanical properties of neutron irradiated Fe–Cr ferritic model alloys with various concentrations of Cr and minor solutes (Ni, Si and P) were investigated along with high-Cr ferritic/martensitic steels. The neutron irradiation was performed in a materials test reactor at two temperatures, 290 ∘C and 450 ∘C, reaching a dose of 0.11 dpa. Tensile tests were subsequently performed on small flat specimens. Significant hardening of about 225 MPa was observed only in model alloys containing Ni, Si and P solute atoms, irrespective of the Cr content and the irradiation temperature. These results suggest that Cr solutes play lesser role in irradiation hardening of ferritic/martensitic alloys steels than previously believed, while minor solutes have a clear effect.