Structural Materials For Fusion Reactors Research Articles

A Kinetic Monte Carlo Model for Helium Diffusion and Clustering in Fusion Environments

Structural materials in fusion reactors will operate in harsh radiation conditions including high displacement rates from 14 MeV neutrons with accompanying high levels of hydrogen and helium production and will experience severe property degradation. Predicting their in-service performance requires a detailed understanding of the mechanisms of defect accumulation and microstructure evolution. The physical processes involved in radiation damage are inherently multiscale, spanning more than 15 orders of magnitude in length and 24 orders of magnitude in time. Here, we describe a kinetic Monte Carlo (KMC) model to simulate the migration and clustering of transmutant helium gas atoms and ultimately determine the role of helium in mediating the long term aging of primary defects (vacancies, self-interstitial atoms and their clusters) produced in displacement cascades. The results illustrate the mechanisms responsible for the formation of vacancy-He clusters, in particular the high mobility of small vacancy-He clusters, which often lead to cluster growth through cluster–cluster coalescence events.

On the potentiality of using ferritic/martensitic steels as structural materials for fusion reactors

Reduced activation ferritic/martensitic (RAFM) steels are the reference structural materials for future fusion reactors. They have proven to be a good alternative to austenitic steels for their higher swelling resistance, lower damage accumulation and improved thermal properties. However, irradiated RAFM steels exhibit a low temperature hardening and an increase in the ductile-to-brittle transition temperature, which imposes a severe restriction on reactor applications at temperatures below about 350°C. Furthermore, a high density of small cavities (voids or helium bubbles) has been recently evidenced in specimens irradiated with a mixed spectrum of neutrons and protons at about 300°C at a dose of 10 dpa, which could affect their fracture properties at intermediate temperatures. The upper temperature for the use of RAFM steels is presently limited by a drop in mechanical strength at about 500°C. New variants that can withstand higher temperatures are currently being developed, mainly using a stable oxide dispersion. This paper reviews European activity in the development of RAFM steels.

Radiochemical reprocessing of vanadium–chromium–titanium alloys after fusion reactor demo decommissioning

Activated structural materials of fusion reactors after the reactor decommissioning may be sent either for burial or for radiochemical reprocessing and consequent material refabrication. This paper considers an opportunity of radiochemical extraction reprocessing of activated vanadium alloy, which is a potential structural material for Russian demonstration fusion reactors (DEMO). The nitrate solutions of the alloy components were used as the aqueous phase. The solution of di-2-ethyl-hexyl-phosphoric acid (D2EHPA) in a hydrocarbon solvent was chosen as an extractant in the experiments. The extraction of alloy impurities (till the dose rate level of 12.5 µSv/h) may be attained in an extractor cascade with 50 stages. End products of the extraction reprocessing are nitrate solutions of vanadium and chromium and titanium hydroxide. The expediency of the radiochemical reprocessing of the alloy or its burial may be determined after the technical and economic comparison of these processes.

High Temperature Compatibility Issues for Fusion Reactor Structural Materials

The primary functions of the blanket in a deuterium/tritium-fueled fusion reactor are to convert neutron kinetic energy into heat, breed tritium for the fuel cycle and accommodate a heat transfer system. Various blanket concepts have been considered including those incorporating liquid metals, molten salts, water and He. The objective of this review is to outline some of the critical compatibility issues associated with structural materials being considered for designs operating at >500°C. Examples are given for vanadium alloys, magnetohydrodynamic coatings, oxide dispersion strengthened ferritic alloys and silicon carbide composites.

Swelling behavior of F82H steel irradiated by triple/dual ion beams

Irradiations for spallation target vessels and structural materials of fusion reactors were simulated using simultaneous triple/dual ion beams consisting of Fe 3+, He + and H + ions or Fe 3+ and He + ions at temperatures between 470 and 600 °C to 50 dpa. The swelling of F82H (Fe–8Cr–2W–0.2V–0.04Ta–0.1C) martensitic steel was enhanced by a synergistic effect of displacement damage and the implantation of helium and hydrogen. The maximum swelling of F82H steel was 3.2% at 470 °C under a simulation of structural materials of fusion reactors, and was higher than 1.2%, which applied to a simulation of spallation target vessel. The swelling under a simulation of fusion reactor decreased with increasing irradiation temperature, however the swelling under a simulation of spallation target vessel was again increased at 600 °C by the high helium concentration. From the microstructural analysis of taking account of cavity growth process, the cause of the enhancement of swelling under a simulation of fusion reactor is thought to be gas pressure of hydrogen and helium in cavities during irradiation. The effects of 50% cold-working and carbon implantation on swelling behavior were also examined. The swelling was reduced from 3.2% to 1.4% by 50% cold-working, and to 0.5% by carbon implantation.

Development of an extensive database of mechanical and physical properties for reduced-activation martensitic steel F82H

Tensile, fracture toughness, creep and fatigue properties and microstructural studies of the reduced-activation martensitic steel F82H (8Cr–2W–0.04Ta–0.1C) before and after irradiation are reported. The design concept used for the development of this alloy is also introduced. A large number of collaborative test results including those generated under the International Energy Agency (IEA) implementing agreements are collected and are used to evaluate the feasibility of using reduced-activation martensitic steels for fusion reactor structural materials, with F82H as one of the reference alloys. All the specimens used in these tests were prepared from plates obtained from 5-ton heats of F82H supplied to all participating laboratories by JAERI. Many of the results have been entered into relational databases with emphasis on traceability of records on how the specimens were prepared from plates and ingots.

Relation between Creep Rupture Strength and Substructure of Heat Resistant Steel.

Creep rupture tests were performed at around 650°C on a low carbon steel, a W-bearing low carbon steel, a low carbon steel containing 0.1 % of TaN and a 0.1 %C-8%Cr-2%W-0.2%V-0.04%Ta steel (F-82H), structural material for fusion reactors. The equivalent obstacle spacing for mobile dislocations is calculated using only rupture data according to the creep theory developed by some of the authors. Thin films taken from the ruptured specimens were examined under a transmission electron microscope (TEM). The observed sub-grain size or the calculated inter-particle distance roughly coincides with the equivalent obstacle spacing. This indicates that the microscopic variable, i.e. the obstacle spacing for mobile dislocations, such as sub-grain size or inter-particle distance, can be directly calculated from only the macroscopic variable, i.e. time to rupture or creep rate.

1127. Ti-6Al-4V接合材料の強度に与える中性子照射効果

In order to investigate applicability of Ti alloy to large scaled structural material for fusion reactors, irradiation effect on the mechanical properties of Ti-6Al-4V alloy and its TIG welded material was investigated after neutron irradiation (temperature: 746-788K, fluence: 2.8×1023n/m2 (>0.18MeV)).The following results were obtained.(1) Irradiated Ti alloy shows about 20-30% increase of its tensile strength and large degradation of facture elongation, comparing with those of unirradiated Ti alloy.(2) TIG welded material behaves as Ti alloy in its tensile test, however, shows 30% increase of area reduction in 373-473K, whereas 1/2 degradation of area reduction over 600K.(3) Irradiated TIG welded material behaves heavier embrittlement than that of irradiated Ti alloy.(4) Charpy impact properties of un- and irradiated Ti alloys shift to ductile from brittle fracture and transition temperature shift, ΔT was estimated as about 100K.(5) Remarkable increase of hardness was found, especially in HAZ of TIG welded material after irradiation.

Benchmark experiment on silicon carbide with D–T neutrons and validation of nuclear data libraries

Silicon carbide (SiC) is one of the low activation structural materials for fusion reactors. A benchmark experiment on SiC was conducted at the deuterium–tritium (D–T) neutron source facility FNS in JAERI, and the validity of the cross section data for SiC was investigated. An experimental assembly made of sintered SiC (457×457 mm, and 711 mm in thickness) was bombarded by D–T neutrons, and neutron spectra, gamma-ray spectra, dosimetry reaction rates and gamma-ray heating rates were measured in the assembly. Computational analyses of the experiment were performed by Monte Carlo transport calculations with the MCNP-4B code and several recent evaluated nuclear data files. The following findings were pointed out. The calculations with JENDL Fusion File and JENDL-3.2 predicted adequately the neutron fluxes from 14 MeV to 0.3 eV as well as the gamma-ray fluxes at any positions up to ∼600 mm depth in the SiC assembly. Results for the calculation with FENDL/E-2.0 were similar to those with the two JENDL calculations, but the neutron fluxes in the energy range from 1 to 14 MeV were calculated slightly smaller. The cross section data for silicon in FENDL/E-1.0 were found inadequate for nuclear design calculations.

Materials research towards a fusion reactor

In this paper the expected goals for the development of structural materials for fusion reactors and important nuclear-specific selection criteria are discussed. A short status of development is given and critical issues as well as necessary investigations are identified for the three major materials groups, the ferritic–martensitic steels, ceramic SiC/SiC composites and vanadium alloys. Proposals for a long-term development strategy are developed, including the use of existing irradiation facilities and an intense 14 MeV neutron source. The International Fusion Materials Irradiation Facility (IFMIF), selected and recommended by the International Fusion Materials Community, is an appropriate and necessary test bed for the materials development.

Structural materials for fusion reactors

In order to preserve the conditions for an environmentally safe machine, at present the selection of materials for the structural components of fusion reactors is made not only on the basis of adequate mechanical properties, behaviour under irradiation, and compatibility with other materials and cooling media, but also on their radiological properties, i.e. radioactivity, decay heat and radiotoxicity. These conditions strongly limit the number of suitable materials to a few families of alloys, generically known as low activation materials. The criteria for making decisions about such materials, the alloys resulting from the application of these ideas and the main issues and problems with their use in a fusion environment are discussed.

Mechanical Characterization of Two Low-Activation Chromium Alloys in As-Received and Heat-Treated Conditions

In recent years, within the fusion long-term programmes, attention has been devoted to the characterization of Chromium (Cr) alloys, in view of their elevated corrosion resistance, low activation properties and high-temperature mechanical strength.As part of the European Fusion Programme, an activity has been launched in 1999 with the aim of exploring the potential of Cr alloys as structural materials in fusion reactors, for example, as first wall or blanket materials. Recent investigations have focused attention on two commercially available materials: high-purity 99.7% Cr (DUCROPUR) and Cr alloyed with 5% Fe and 1%Y203 (DUCROLLOY), both of which have shown excellent low activation characteristics.The mechanical properties of these two alloys, in both as-received and heat-treated conditions, have been characterized at SCK•CEN by means of tensile, instrumented impact and static three-point bend tests, using standard and sub-size specimens. Tensile tests have also been carried out on samples irradiated at 300 °C in the BR2 reactor in Mol up to an accumulated dose of about 0.5 dpa.

Ductility correlations between shear punch and uniaxial tensile test data

The shear punch test was developed to address the need of the fusion reactor structural materials community for small scale mechanical properties tests. It has been demonstrated that effective shear strength data obtained from the shear punch test can be linearly related to uniaxial tensile strength for a wide variety of alloys. The current work explores the existence of a similar relationship between shear punch test data and both the tensile strain hardening exponent and the uniform elongation.

Critical issues and current status of SiC/SiC composites for fusion

SiC fiber-reinforced SiC matrix composites (SiC/SiC) are being considered as a candidate structural material for fusion reactors because of their low induced radioactivity by 14 MeV neutron irradiation and high-temperature strength. Material development of SiC/SiC composites, including SiC fiber, matrix and interphase processing for fusion reactor applications, has been under investigation for several years. The purpose of this paper is to present an overview of current research in material development, properties and irradiation response of new SiC/SiC composites. Several technologically critical issues to be solved for fusion applications, such as irradiation resistance, thermal properties, environmental effects, hermeticity, joining technique and protective coating are also reviewed.

Irradiation creep at 60°C in SUS 316 and its impact on fatigue fracture

Structural materials in fusion reactors will be subjected to irradiation by energetic particles at temperatures widely ranging from liquid He to above 1000°C. Hence, the very large irradiation creep strain at 60°C previously reported in the ORR/ORNL pressurized tube experiment to 8 dpa is important. Computer calculations for the 20% cold-worked SUS 316 demonstrated the transient nature of this radiation-induced creep, caused by the overwhelming flux of excess interstitial atoms lasting nearly one year at 60°C where the diffusivity of vacancies is very low. In order to confirm such a transient nature, continuous creep measurement under irradiation is necessary and was carried out using 17 MeV protons. Development of very significant creep strain, much larger than that at 300°C, and steadily decreasing creep rate were observed at 60°C as the calculation predicted. A significant influence of the dynamic irradiation effect at 60°C on fatigue fracture was also observed.

High-performance SiC/SiC composites by improved PIP processing with new precursor polymers

As they are potential candidates for fusion reactor structural materials, R&D are being conducted on SiC-based composite materials (CREST-ACE program). To improve the efficiency of the polymer impregnation and pyrolysis (PIP) process for SiC/SiC composite fabrication, a new precursor polymer, poly-vinylsilane (PVS) with SiC filler addition, was adopted as a matrix precursor and process optimization was performed. Consequently, high-density SiC/SiC composite with high mechanical properties was efficiently fabricated. Importantly, near-stoichiometric SiC matrix was developed by blending of poly-carbosilane (PCS) and poly-methylsilane (PMS). Remarkable improvements in tensile properties and fatigue characteristics (at 1573 K) were attained when inorganic powder fillers, BMAS or ZrSiO4, were added to the polymer mixture as the matrix precursor. These results are encouraging to make economically and environmentally attractive fusion reactors utilizing SiC/SiC composites as major structural materials.

Long-term activation potential of the steel eurofer as structural material of a demo breeder blanket

The long-term activation behavior of the martensitic–ferritic steel Eurofer 97 as structural material for fusion reactors has been investigated on the basis of activation calculations for the European Union (EU) helium-cooled pebble bed (HCPB) outboard DEMO blanket. Activation calculations were performed with the FISPACT-97 inventory code and EAF-97 activation cross-section data for three representative positions at the outboard first wall, the central part and the back wall of the EU-HCPB DEMO blanket. All stable elements were taken into account as impurity elements. Radial dependent values and blanket averages were derived for the permissible impurity element concentrations to arrive at an Eurofer 97 gamma dose rate below the hands-on level after a cooling time of 200 years.

Experimental investigation of radioactivities induced in fusion reactor structural materials by 14-MeV neutrons

The aim of the present study is the experimental validation of European activation system (EASY-97) for structural materials irradiated with 14-MeV neutrons. Samples of the stainless steel 316 (AISI 316 LN (IG)), which is the structural material of international thermonuclear experimental reactor (ITER), of the low-activation steels MANET and F82H and of the vanadium alloys V–3Ti–lSi, V–4Ti–4Cr and V–5Ti–2Cr were used for benchmarks. Pieces having the mass of ≈l g were irradiated at the high-intensity neutron generator SNEG-13 at Sergiev Posad with neutrons of 14.93 and 14.37 MeV mean energy. Fluences applied were of the order of 10 14 cm −2. γ-Spectra of the samples were measured several times during the decay up to about 1-year. The activities of 27 different radionuclides, identified by γ-energies and half-life, were determined. Their half-lives range from a few min up to 5.3 years. For each of the measured values the activity was calculated with EASY-97, and calculated-to-experimental values ( C/ E) were determined. If the uncertainty of the C/ E is estimated with both the experimental errors and the calculational one, including cross-section as well as half-life uncertainties, validation of EASY-97 can be stated for most of the activities. The pathways of production are analysed for all activities, and the corresponding cross-sections are discussed if larger deviations were observed.

Operating temperature windows for fusion reactor structural materials

A critical analysis is presented of the operating temperature windows for nine candidate fusion reactor structural materials: four reduced-activation structural materials (oxide-dispersion-strengthened and ferritic/martensitic steels containing 8–12%Cr, V–4Cr–4Ti, and SiC/SiC composites), copper-base alloys (CuNiBe), tantalum-base alloys (e.g. Ta–8W–2Hf), niobium alloys (Nb–1Zr), and molybdenum and tungsten alloys. The results are compared with the operating temperature limits for Type 316 austenitic stainless steel. Several factors define the allowable operating temperature window for structural alloys in a fusion reactor. The lower operating temperature limit in all body-centered cubic (BCC) and most face-centered cubic (FCC) alloys is determined by radiation embrittlement (decrease in fracture toughness), which is generally most pronounced for irradiation temperatures below ∼0.3 T M where T M is the melting temperature. The lower operating temperature limit for SiC/SiC composites will likely be determined by radiation-induced thermal conductivity degradation, which becomes more pronounced in ceramics with decreasing temperature. The upper operating temperature limit of structural materials is determined by one of four factors, all of which become more pronounced with increasing exposure time: (1) thermal creep (grain boundary sliding or matrix diffusional creep); (2) high temperature He embrittlement of grain boundaries; (3) cavity swelling (particularly important for SiC and Cu alloys); or (4) coolant compatibility/corrosion issues. In many cases, the upper temperature limit will be determined by coolant corrosion/compatibility rather than by thermal creep or radiation effects. The compatibility of the structural materials with Li, Pb–Li, Sn–Li, He and Flibe (Li 2BeF 4) coolants is summarized.

1086.Fe-Mn-Cr系低放射化高強度非磁性鋼の開発(I)

It is very important to develop low activated/non-magnetic materials as large scale structural materials for fusion reactors. In the structural design of JT-60SU, low activated/non-magnetic materials with high specific strength and low decay heat characterizations are required. In the present paper, a new low activated/non-magnetic material (15.5Mn-16Cr-0.2N-0.3Si-0.2C (wt%)) based on the conventional high manganese steel with lower Ni, Co, C, N and Mn contents for the purpose of lower activation and decay heat was developed and the mass production conditions of the material were optimized. Fundamental material characterization tests of the new material developed in present study were carried out, and the following conclusions are derived; (1) Lower activation characterizations with the new materials in the order of 1/10 of that of SUS316L steel (2) Higher strength of the material in the order of 2-3 of SUS316L steel (3) Lower decay heat with higher thermal conductivity with comparison of SUS316L steel