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Tungsten potassium (WK) alloy has been reported as one of the ideal plasma-facing materials (PFMs). Tantalum alloying is a good method to improve the mechanical properties of tungsten. The effect of tantalum contents on the irradiation resistance of WK alloy has not yet been reported. In this study, WK (containing 82 ppm potassium) alloy with 1 wt. % Ta and 3 wt. % Ta, specifically WK-1Ta and WK-3Ta, were fabricated with sparking plasma sintering and irradiated with 7.5 MeV W2+ ion. The relative densities of WK-1Ta and WK-3Ta are 97.2% and 96.4%, respectively. The average grain sizes of WK-1Ta and WK-3Ta are 2.08 μm and 1.51 μm, respectively. The Vickers hardness of WK-3Ta is nearly 20% higher than that of WK-1Ta, both before and after irradiation. Irradiation hardening was confirmed by nano indentation test results. After irradiation, the number of dislocation loops formed in WK-1Ta and WK-3Ta are very similar, and the dislocation loop density of WK-3Ta is only slightly higher than that of WK-1Ta. This phenomenon is consistent with nano hardness analysis results. Compared to the reported nano hardness results of WK alloys, both WK-1Ta and WK-3Ta had higher hardness than the WK alloys before irradiation. Compared to the irradiation hardening results for the reported WK alloys, the existence of Ta may have positive influence on resistance to irradiation hardening.

Tungsten fiber-reinforced tungsten composite (Wf/W) material is considered a plasma-facing material (PFM) with good application prospects. Commercial tungsten wire (fiber) prepared through forging and drawing processes has excellent mechanical properties, as well as a very high recrystallization temperature due to the unique texture of it grain structure. Commercial tungsten fiber is the most proper reinforcement for Wf/W. The change in the properties of tungsten fiber because of neutron irradiation makes it inevitable for Wf/W to be used as PFMs. However, there is very little research on the change in the properties of tungsten fiber caused by neutron irradiation. In this work, we used heavy ion irradiation to simulate the displacement damage generated by neutron irradiation to explore the alteration of the properties of a commercial tungsten fiber caused by neutron irradiation. The investigated subject was tungsten fiber with a diameter of 300 μm. The irradiation source was 7.5 MeV Au2+, which generated a maximum displacement damage of 60 dpa at a depth of 400 nm, and the irradiation influenced depth was 1000 nm. Because of the irradiation, significant lattice distortion occurred within the tungsten fiber, resulting in the transition from (110) texture to (100) texture at the fiber’s cross-section. The results of the Schmidt factor and Taylor factor analysis indicate a decrease in the plasticity of the tungsten fiber after irradiation, but it did not completely lose its plasticity. The results of the nanoindentation test confirmed the radiation hardening. After irradiation, the hardness of the tungsten fiber increased by approximately 0.33 GPa, but this increase was relatively small compared to other tungsten-based materials. This indicates that commercial tungsten fiber is a low-cost and highly reliable reinforcement material for Wf/W composite materials.

Blanket, including its first wall (FW), is one of the key components in the fusion reactor, which works in the severe condition with the huge high energy neutron irradiation, high heat flux and electromagnetic load. Facing these technical challenges for the blanket and its FW, Southwestern Institute of Physics is engaging in the development of the blanket technology following the proposed China magnetic confinement fusion development strategy. This paper presents the design and R&D progress of the ITER FW and blanket, Chinese helium cooled ceramic breeder test blanket module (HCCB TBM) for ITER and HCCB tritium blanket for China Fusion Engineering Test Reactor (CFETR). The design of the ITER enhanced heat flux FW has been updated to improve manufacturability and to reduce water leak risk. Key technologies have been qualified by high heat flux tests of small mock-ups and full-scale plasma-facing fingers. New assembly technologies are under development. Following the completion of full-scale prototype qualification, the ITER shielding blanket modules are under manufacturing for series production. The preliminary design of HCCB TBM for ITER is under implementation to improve the manufacturability and inspectability based on the fabrication study. Two prototypes of TBM submodule have been fabricated and tested to validate the feasibility of manufacture procedure, which are finally assembling into one semi-prototype of TBM. The preliminary design of HCCB tritium blanket for CFETR has been completed and the R&D has been started, especially for the special technologies different with the ITER blanket and TBM. Meanwhile, the progress of reduced activation ferritic/martensitic steel and function materials for blanket are also presented.

In this study, CuCrZr/316L(N)-IG explosion bonding bimetallic plates were produced and conducted the same thermal cycles as the ITER enhanced heat flux (EHF) First Wall (FW) fingers. The bonding in different states was investigated using optical and scanning electron microscopy and tensile, hardness tests. Wavy morphology and deformation structure were observed in the as-received explosion bonding plates. After solution annealing, deformation feature disappeared and recrystallization of CuCrZr alloy happened. The plates can meet the design requirement for the EHF FW components after the following Be/CuCrZr HIPing, both for the microstructure and strength.

W and RAFM-steel joint is the major feature of the plasma facing component in the first wall of a fusion reactor. In present work, W and RAFM-steel were successfully diffusion bonded with an Fe interlayer by hot isostatic pressing at 980 °C and 150 MPa for 1 h. Two Fe-rich phases, a Fe2W type phase and a Fe3W3C type phase, were formed at the W/Fe interface. Both phases were drastically deformed under the interfacial thermal stress, leading to stress-induced martensitic transition of the Fe2W type phase and plastic slip of the Fe3W3C type phase. Their deformation could reduce the residual stress significantly at the W/reaction layer interface and the W/Fe/RAFM-steel joint exhibited high bonding strength (>240 MPa) and ductility in the shear test.

In this study, a 9Cr oxide-dispersion strengthened (ODS) alloy with additional corrosion resistant element Al was fabricated by mechanical alloying (MA) and hot pressing (HP) to explore the impact of Al on the microstructure and mechanical property of a 9Cr ODS alloy. It is found that the Al completely dissolved into the Fe–Cr matrix after milling for 30 h. The minor phases in the Al-containing 9Cr ODS ferritic alloy were investigated by a high-energy X-ray, and were identified to be orthorhombic- YAlO 3 (YAP), bcc- Y 3 Al 5 O 12 (YAG), monoclinic- Al 2 Y 4 O 9 (YAM), and hexagonal- YAlO 3 (YAH). These phases were further confirmed by selected area diffraction pattern (SADP), energy dispersive spectroscopy (EDS), and high resolution transmission electron microscopy (HRTEM). In addition, their volume fractions were also calculated from the integrated intensities. According to the analysis of the particles and their formation sequences, the larger particles (greater than 100 nm) are identified as mainly YAG and Al 2 O 3 particles, while the particles with small size (less than 30 nm) are likely primarily YAM, YAH, and YAP particles. The yielding strength (YS) and ultimate tensile strength (UTS) at RT are 563 MPa and 744 MPa, respectively, while the YS and UTS at 700 °C are 245 MPa and 276 MPa, respectively. Although the addition Al in ODS alloys decreases the strength at RT, the values at high temperature are similar to those obtained for 9Cr ODS alloys strengthened by fine Y–Ti–O particles.

ITER will be used to test tritium breeding module concepts, which will lead to the design of DEMO fusion reactor demonstrating tritium self-sufficiency and the extraction of high grade heat for electricity production. China plans to test the HCCB TBM modules during different operation phases. Related design and R&D activities for each TBM module with the auxiliary system are introduced. The helium-cooled ceramic breeder (HCCB) test blanket module (TBM) is the primary option of the Chinese TBM program. The preliminary conceptual design of CN HCCB TBM has been completed. A modified design to reduce the RAFM material mass to 1.3 ton has been carried out based on the ITER technical requirement. Basic characteristics and main design parameters of CN HCCB TBM are introduced briefly. The mock-up fabrication and component tests for Chinese test blanket module are being developed. Recent status of the components of CN HCCB TBM and fabrication technology development are also reported. The neutron multiplier Be pebbles, tritium breeder Li 4 SiO 4 pebbles, and structure material CLF-1 of ton-class are being prepared in laboratory scale. The fabrication of pebble bed container and experiment of tritium breeder pebble bed will be started soon. The fabrication technology development is proceeding as the large-scale mock-up fabrication enters into the R&D stage and demonstration tests toward TBM testing on ITER test port are being done as scheduled.

The thermal diffusivity of reduced activation ferritic/martensitic steel (CLF-1), which is recognized as the primary candidate structural material for the test blanket module of the international thermal-nuclear experimental reactor, has been studied by the time-domain (TD) photoacoustic piezoelectric (PAPE) technique. The TD PAPE model based on a simplified thermoelastic theory under square-wave modulated laser excitation is presented, relating the TD PAPE signal to the modulation frequency, thermal diffusivity, and other material parameters. Thermal diffusivities of reference samples such as copper and nickel were measured and analyzed, by which the validity of the technique is verified. The thermal diffusivity of the CLF-1 sample was measured to be $$8.2\,\text {mm}^{2}{\cdot }\text {s}^{-1}$$ , which is at a medium level among the ordinary steel materials ( $$3\,\text {mm}^{2}{\cdot }\text {s}^{-1}$$ to $$14\,\text {mm}^{2}{\cdot }\text {s}^{-1})$$ and has decent heat-dissipation ability. The results show that the TD PAPE technique can provide a fast and economic way for the investigation of the thermophysical properties of fusion reactor structural materials.

In this work, thermal ageing of the reduced activation ferritic/martensitic steel CLF-1 with a nitrogen content of 0.025% has been performed at temperatures of 823 K and 873 K for 6000 h. Mechanical properties (hardness, tensile and Charpy impact) and microstructure characterisation has been carried out. After thermal ageing treatments, the steels showed similar values of hardness and tensile properties and a slight deterioration of impact properties were observed. After long term thermal ageing, the M23C6-type precipitation coarsens and agglomerates along lath and grain boundaries. Compared with steels that have lower nitrogen content, the coarsening and agglomeration of M23C6 in the steels with higher nitrogen content is much lower and the steel shows higher microstructural stability. (C) 2013 Elsevier B. V. All rights reserved.

Reduced activation ferritic/martensitic steel (RAFM) is recognized as the primary candidate structural material for ITER's test blanket module (TBM). To provide a material and property database for the design and fabrication of the Chinese helium cooled ceramic breeding TBM (CN HCCB TBM), a type of RAFM steel named CLF-1 was developed and characterized at the Southwestern Institute of Physics (SWIP), China. In this paper, the R&D status of CLF-1 steel and the technical issues in using CLF-1 steel to manufacture CN HCCB TBM were reviewed, including the steel manufacture and different welding technologies. Several kinds of property data have been obtained for its application to the design of the ITER TBM.

Current progress on the design and R&D of Chinese helium-cooled solid breeder test blanket module, CN HCSB TBM is presented. The updated design on structural, neutronics, thermal-hydraulics and safety analysis has been completed. In order to accommodate the HCSB TBM ancillary system, the design and necessary R&Ds corresponding sub-systems have being developed. Current status on the development of function materials, structure material and the helium test loop are also presented. The Chinese low-activation ferritic/martensitic steels CLF-1, which is the structural material for the HCSB TBM is being manufactured by industry. The neutron multiplier Be and tritium breeder Li4SiO4 pebbles are being prepared in laboratory scale.

The reduced activation ferritic/martenstic steel CLF-1 prepared by the Southwestern Institute of Physics in China was irradiated by helium ions with an energy of 5 keV at room temperature using an electron cyclotron resonance (ECR) ion irradiation apparatus. After the irradiation, the helium retention and desorption were investigated using a technique of thermal desorption spectroscopy (TDS). The experiment was conducted with both the normal and welded samples. Blisters were observed after the helium ion irradiation, and the surface density of blisters in the welded samples was lower than that in the non-welded samples. Three desorption peaks were observed in both the non-welded and welded samples. These desorption peaks corresponded to those of blister ruptures and the helium release from the inner bubbles and the defects. The amount of helium retained in the welded samples was approximately the same as that in the non-welded samples, which was much less than other reduced activation materials, such as vanadium alloy and SiC/SiC composites.

Testing TBM (Test Blanket Modules) is one of important engineering test objectives in ITER project. China is performing the TBM design and R&D based on Chinese development strategy of fusion DEMO. Helium-cooled test blanket module concept with ceramic breeder for testing on ITER will be one of the basic options in China. The current progress and status on design and R&D of CH HC-SB (Chinese Helium-cooled Solid Breeder) TBM are introduced briefly in this report. The modified designs of HC-SB TBM, related ancillary sub-systems and test strategy on ITER as well as relevant R&D activities are summarized.

Chinese helium-cooled ceramics breeder test blanket module (CH HCCB TBM) is determined to be tested in ITER machine to get data for fusion reactor design and development in future. Chinese TBM is designed to occupy half of port C with 484mm in torroidal and 1660mm in poloidal. Radial length is 675mm. TBM is composed of box, 12 submodules and independent backplate. Box formed by first wall, grids and caps have 12 caivities to hold submodules. Box and submodules are supported by backplate by welding. Backplate distribute helium with flow rate 1.36kg/s to cool first wall and then part of it go out of TBM by bypass. The rest 0.77kg/s go on to cool caps and girds first and then cool submodules. Submodules with dimensions 250mm×202mm×318mm have independent cooling and purging systems connected to backplate manifold systems. In a submodule, two U-shaped structures hold breeding material Li4SiO4 pebbles. Out of the structure filled beryllium pebbles. Neutronics results show that tritium production is ∼64mg/FPD. Maximum temperature 538°C of structure material occurs in the front of first wall with surface heat flux 0.5MW/m2. Maximum total stress at first wall is 471MPa at 394°C; that in submodules is 426MPa at 400°C; that in backplate is 526MPa at 410°C, In order to explore development technologies for the TBM, a mockup with dimensions 484mm (torroidal)×592mm (poloidal)×675mm (radial) has been designed. The mockup with similar structure ignores bypass and purge gas systems. In the mockup, there’s only one submodule and the other three are replaced by submodule replacements. By discussions and investigations, development route has been decided and the mockup is being fabricated.Copyright © 2013 by ASME