Papers by Keyword: Fusion Reactor

Abstract: Within the large scale fusion experimental device ITER, Ion Cyclotron Resonance Heating (ICRH) system is one of the three heating systems which will supply total heating power of 20 MW (40-55 MHz) up to one hour operation. Radio-Frequency (RF) contacts are integrated within the antennas for assembly and operation considerations, which will face extremely harsh service conditions, including neutron irradiation, heavy electrical loads (RF current reaches up to 2 kA with a linear current density of 4.8 kA/m), high thermal loads and also long-duration vacuum baking at 250°C before each experimental plasma campaign. CuCrZr and 316L steel have been shown to be proper base candidate materials for ITER RF contact louvers and conductors respectively. However, in order to limit the wear and the diffusion phenomena at the RF contact as well as to reduce the contact resistance, functional protective layers should be developed. The aim of this work is to investigate Au-Ni and Rh functional layers, electroplated on CuCrZr and 316L respectively. The efficiency of the Au-Ni/Rh coated pairs was evaluated through thermal ageing diffusion tests, using EDS cross-section mapping and XRD techniques. Wear and electrical contact performances of the Au-Ni/Rh original and thermally aged pairs have also been deeply studied on a dedicated tribometer operated at ITER relevant conditions.

Abstract: National Institute for Fusion Science (NIFS) launched in 2014 a research program for developing Dispersion Strengthened (DS) Cu alloys for application to the heat sink materials of divertors of fusion reactors, using newly installed ball-milling, encapsulation, and Hot Isostatic Pressing (HIP) facilities. A unique feature of these facilities is that the entire process can be performed without exposing the materials to air, enabling precise impurity control. Cu-Al, Cu-Zr and Cu-Y alloys have been produced in this program. Various technological advancement has been made for the fabrication, such as suppression of powder adhesion to the wall of containers during the ball milling, and encapsulation technology including development of small volume tubular capsules.

Abstract: A fusion reactor is expected as one of the new electric power sources in next generation. Reduced activation ferritic/martensitic steel F82H is planned to be used as a structural material for the blanket modules set on the inner wall of the reactor. However, especially in the case of laser beam welding (LBW), the weldability of the steel was not completely clarified. On the other hand, although post weld heat treatment (PWHT) should be conducted for the welds of the steel in accordance with general standards for chrome steels, the heat treatment conditions were uncertain. Therefore, adaptability of LBW as a joining method for the steel and the applicable PWHT conditions for the welded joints were investigated in this study. The effect of LBW conditions on weld penetration behavior were ascertained by observation of cross sections in the welds. The adequate PWHT conditions were confirmed in consideration of both hardness distributions measured in welds and ductile-brittle transition temperatures (DBTT) evaluated using Charpy impact test. Full penetration without weld defects such as hot cracking, porosity etc. was obtained for plates with the thickness of 4mm of the steel by control welding conditions. That means laser beam is one of useful welding heat sources to realize sound weld joints of the steel. In addition, due to select appropriate PWHT conditions, the hardness in welds was suppressed to the level of base metal and the toughness in the welded joints was improved to a practical level without the damage to base metal.

Abstract: In-vessel tritium inventory in fusion reactors has to be reduced from a view point of safety in fusion reactors. It is required to evaluate the amount of tritium retained in tungsten plasma facing walls. The plasma discharge with hydrogen isotope (deuterium) was conducted to evaluate the tritium retention in tungsten. The glow discharges using helium, neon and argon were performed after the deuterium discharge to reduce the deuterium retention. The use of inert gas discharge little reduced the deuterium retention. Namely, the inert gas glow discharge is not useful to reduce the tritium inventory. The deuterium glow discharge significantly replaced the hydrogen in the tungsten wall into the deuterium. Thus, the deuterium glow discharge is quite useful to reduce the tritium inventory through the hydrogen isotope exchange. The use of neon or argon glow discharge followed by deuterium discharge can more reduce the tritium inventory. In addition, the tritium inventory can be easily reduced if the wall baking with a temperature of 700-800K is conducted.

Abstract: Nuclear fusion is one of the most important ways to resolve human energy issues. The main structure of nuclear fusion device and its functions are introduced in this article. Virtual simulation technology has been used in the process of the conceptual design, detailed design and optimization of the fusion device structure. Electromagnetic analysis of the ITER feeder system show the specific application of virtual simulation technology in nuclear fusion engineering, as well as the thermal analysis of ITER thermal shield. It verifies the importance of virtual simulation technology.

Abstract: The realization of a fusion reactor is critically dependent on the successful development of high performance materials. Especially, the plasma facing components (PFCs) which basically consist of a direct plasma facing armor material and a heat sinking material. Tungsten (W) and Copper-alloy (CuCrZr) have been considered as the potential candidates for armor materials and heat sinking materials, respectively, due to their attractive nuclear and physical properties. However, due to the incompatibility of the coefficient of thermal expansion and the elastic properties between the W and the Cu-alloy as well as the non-homogeneous temperature distribution in PFCs, one of the crucial issues is the generation of thermally-induced residual stresses in W/CuCrZr PFC on cooling either during fabrication or during operation of fusion reactor. Therefore, the thermo-mechanical response of PFCs under high heat flux from the fusion reactor is a critical issue for the development of fusion technology. In the present work, in order to optimize the thermal and mechanical integrity of PFCs, thermally-induced residual stresses in W/CuCrZr PFCs with a compliant interlayer (OFHC-Cu: Oxygen Free High Conductivity Copper) are analyzed numerically by means of finite element method. Result indicated that the use of interlayer in PFCs could significantly reduce the magnitude and the concentration of thermally-induced stresses in comparison to the PFCs without interlayer. And also the optimum thickness for interlayer was suggested based on the current analysis conditions.

Abstract: Ferritic/Martensitic steels, with chromium contents ranging between 9 and 12%, were introduced into fusion material programs due to their better creep resistance and excellent thermal and nuclear properties compared to austenitic stainless steels. Reduced activation ferritic/martensitic (RAFM) steels are considered promising candidates for the test blanket modules of the future International Thermonuclear Experimental Reactor (ITER), being EUROFER steel is the EU reference material. It is a 9 % Cr RAFM steel which exhibits a tempered martensitic microstructure and presently allows operation up to 550 ⁰C. This paper shows the work carried out to develop at a pilot plant scale a Reduced Activation Ferritic/Martensitic (RAFM) steel, Asturfer ®, with chemical composition and mechanical properties very close to EUROFER steel.

Abstract: Different coating technologies, such as plasma spray (PS), physical vapor deposition (PVD) and chemical vapor deposition (CVD), which can fabricate the PFM and join it to heat sink materials simultaneously, were applied for the fabrication of plasma facing materials (PFM) in fusion reactor. In the Institute of Nuclear Materials, University of Science and Technology Beijing (USTB), the concept of functionally graded materials (FGM) was adopted to fabricate coatings for effectively alleviating the thermal stress generated between coatings and the substrate materials under high heat flux loading (5~20 MW/m2). In the last several years, functionally graded coatings, including B4C/Cu, W/Cu and Mo/Cu systems were successfully fabricated by atmospheric plasma spray (APS). Characterization of coatings was performed in order to assess microstructure, mechanical properties and high heat flux properties of the FGM coatings. Furthermore, a high thick tungsten coating with 4 mm on copper – chromium - zirconium (Cu, Cr, Zr) alloy substrates was fabricated by APS. The porosity of the coating is less than 2% while mean tensile strength of the coating is about 7 MPa. However, the content of oxygen in the coating is about 6 wt% by energy dispersive spectrum (EDS) analysis, thus further optimization is necessary.

Abstract: The oxide dispersion strengthened (ODS) ferritic steel and non-ODS reduced-activation ferritic (RAF) steel were irradiated at 773 K by means of a dual-beam ion irradiation technique to a dose of 0.4 dpa with simultaneous helium implantation up to 1000 appm. Microstructural changes were investigated by transmission electron microscopy. The RAF steel showed a preferential formation of cavities at grain boundaries, precipitate interfaces and dislocations. In contrast, the ODS ferritic steel showed a homogeneous and fine distribution of cavities in the matrix. This paper discusses the superior resistance of the ODS ferritic steel against development of cavities in terms of the effects of nano-oxide particles dispersed in the matrix.

Abstract: The recently developed small punch (SP) creep test was applied to four different heatresistant ferritic steels, namely, two kinds of conventional ferritic steels which had been actually used in the high-temperature components for long periods and two advanced high chromium ferritic steels for fusion reactor materials to investigate the applicability of the SP creep test. The ratio of the load of SP creep test to the stress of standard uniaxial creep test was calculated so that both the creep rupture curves (load/stress versus Larson-Miller parameter curves) were overlapped to convert the results of SP creep test into those of standard test. As a result, the ratio was determined to be 2.4, irrespective of the kind of ferritic steel. This result indicates that the creep rupture strength of heat-resistance ferritic steels can be estimated using a miniaturized plate-type specimen and this conversion coefficient 2.4 independent of the kind of ferritic steel.