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The development of Tokamak fusion reactors encompasses hybrid and pure fusion reactor designs. In the study of tritium self-sustainability, the average retention time method is predominantly employed for Tokamak fusion reactors, with less emphasis on hybrid reactors. To enhance research in this area, this paper employs a more realistic integral analysis method to examine the requirements for startup tritium inventory and tritium breeding ratio(TBR) necessary to achieve tritium self-sustainability in both hybrid and pure fusion reactors. The findings demonstrate a linear relationship between startup tritium inventory, backup tritium inventory, and fusion power, while the required TBR is inversely proportional to the fusion power. Notably, the reserve tritium inventory plays a significant role in determining startup requirements for hybrid reactors. Within the range of 50-450 MW fusion power, a TBR above 1.15 is needed; particularly below 100 MW fusion power level where it reaches 1.4-posing substantial engineering challenges. The impact of initial levels of startup tritium on required TBR in hybrid reactors is negligible; however, reducing long-term tritium retention can help lower TBR requirements. For pure fusion reactors as well, it is observed that reserve tritium constitutes most of the startup inventory with tens of kilograms being necessary; therefore, considering redundant design or improving maintainability and reliability should be explored to reduce reserve tritium. Furthermore, achieving a required TBR below 1.15 appears more feasible within a range of 1-5 GW fusion power for pure fusion reactors. This paper further examines the impact of operational factors on the required TBR, and the study demonstrates that operational factors are an essential prerequisite for achieving tritium self-sustainability. To emphasize the influence of operational factors on tritium self-sustainability, this paper proposes a redefined relational equation for assessing tritium self-sustainability in fusion reactors.

In terms of wettability, active systems are characterized by a reduction in interfacial energy as the time at specific conditions is increased. This article aims to investigate the evolution of wettability and microstructure, which undergoes a critical transformation at temperatures and dwell times near brazing conditions due to their significant impact on resultant mechanical properties. The objective is to enhance wettability and prevent the formation of different phases that can occur rapidly within the brazing window conditions. Up to 1105 °C, complete fusion of the filler does not occur. However, once it happens, the expansion of the copper filler in EUROFER increases up to 400%, and the contact angle reduces from 100° to 10°, indicating an active wetting behavior. On the other hand, when copper is used with tungsten, an inert behavior is observed, maintaining the contact angle around 70°. Brazed joints carried out under the most promising wetting conditions demonstrated that at 1110 °C-1 min, various phenomena began to occur. This includes solid-state diffusion of copper in the EUROFER, following the austenitic grain boundaries, and partial dissolution of Fe in the copper braze. Increasing the brazing time from 2 to 5 min achieved high interfacial adhesion properties and controlled the diffusion layer and Fe-rich band formed at the W-braze interface, resulting in the best mechanical results (295 MPa). [ABSTRACT FROM AUTHOR]

The divertor must simultaneously withstand unprecedented high heat fluxes of up to 20 MW/m2 and high-energy neutron irradiation of 14-MeV during fusion reactor operation. Accordingly, it is necessary to simultaneously meet the excellent heat dissipation capacity of the divertor and maintain good material performance in harsh neutron irradiation environments. The flat-type divertor demonstrates better heat transfer performance compared to monoblock divertor. Nevertheless, under the condition of a heat flux of 20 MW/m2, the heat transfer and thermal fatigue performance of flat-type divertor using advanced materials are still unidentified. In this study, we conducted a thorough analysis of the heat transfer capabilities and thermal fatigue characteristics of potassium-doped tungsten (KW)/Cu/Oxide dispersion strengthened copper (ODS-Cu)/reduced activation ferritic/martensitic (RAFM) divertor mockup adopts hypervaportron (HV) structure by numerical simulations combined with experiments. The numerical results indicate that the flat-type KW/Cu/ODS-Cu/RAFM divertor mockup expresses excellent heat transfer capacity. The contact area between the edge of the fins and the bottom surface of the ODS-Cu heat sink induces a significant increase in water flow velocity to ∼15 m/s. The peak temperature of loaded KW surface is only ∼985 °C under the flow rate of 5 t/h and inlet temperature of 20 °C. During the high heat flux tests, the prepared flat-type divertor mockup successful endured 1000 cycles of 20 MW/m2 with the peak temperature of 883 °C, and the surface temperature experienced a fluctuation of 2.4 % during the thermal fatigue tests. This study can provide a strong data reference and technical support for the development of fusion reactors, and is of great significance in advancing the commercialization of fusion energy.

Li4SiO4 and Li2TiO3 are granular materials in the pebble bed of fusion reactors, and their physical parameters and mechanical properties directly affect the working status and structural design of the pebble bed. The force chain can be employed to intuitively describe the mechanical properties of the pebble bed in the particle aggregate. Based on the discrete element method, we conducted a numerical study on the evolution characteristics of the force chain of Li4SiO4 and Li2TiO3 pebble beds under biaxial compression and explored the friction coefficient and pebble bed effect of the aspect ratio on the force chain distribution. The results revealed that as the particle friction coefficient increased, the force chain tended to be isotropically distributed, indicating that the friction mechanism was more conducive to a uniform distribution of the force chain. During compression, the average coordination number of the pebble bed increased with the friction coefficient. The Li2TiO3 particles had larger gravity than the Li4SiO4 pebble bed therefore, the Li2TiO3 pebble bed accounted for more force chains in the vertical direction. When the aspect ratio of the pebble bed was less than 0.5 or greater than 2.5, the distribution of force chains exhibited strong anisotropy. Conversely, when the pebble bed aspect ratio ranged between 0.5 and 2.5, the distribution of force chains tended towards an isotropic trend. Moreover, the number of force chains in each direction varied with changes in the aspect ratio of the pebble bed, characterized by a high concentration at the edges and a lower concentration in the middle. The results can provide an in-depth understanding of the force chain distribution and evolution characteristics in the pebble bed and provide a theoretical basis for designing and analyzing tritium breeding pebble beds.

In this paper, a cable-driven endoscopic manipulator (CEM) is designed for the Chinese latest compact fusion reactor. The whole CEM arm is more than 3000 mm long and includes end vision tools, an endoscopic manipulator/control system, a feeding system, a drag chain system, support systems, a neutron shield door, etc. It can cover a range of ±45° of the vacuum chamber by working in a wrap-around mode, etc., to meet the need for observation at any position and angle. By placing all drive motors in the end drive box via a cable drive, cooling, and radiation protection of the entire robot can be facilitated. To address the CEM motion control problem, a discrete trajectory tracking method is proposed. By restricting each joint of the CEM to the target curve through segmental fitting, the trajectory tracking control is completed. To avoid the joint rotation angle overrun, a joint limit rotation angle optimization method is proposed based on the equivalent rod length principle. Finally, the CEM simulation system is established. The rationality of the structure design and the effectiveness of the motion control algorithm are verified by the simulation.

We conducted absorption experiments on Ti using hydrogen isotopes, protium and deuterium, to gain insights into the kinetics of hydrogen atoms, particularly across the Ti surface. Rather than relying on various surface pretreatments, we identified the initial temperature of Ti in vacuum as the determining factor in achieving a high absorption rate and content. Specifically, we found that the initial Ti temperatures of approximately 900 and 980 °C are optimal for protium and deuterium, respectively, likely due to sufficient removal of the native surface oxide on Ti. The dependence of the absorption rate and amount of hydrogen isotopes on the temperature during absorption was partially explained by the phase transformation and exothermicity of the hydrogenation reaction. Moreover, this dependence strongly indicated the existence of a tunneling transport process. Additionally, we developed a kinetic model for hydrogen isotope transport in Ti to conduct numerical simulations. The absorption curves, calculated using the developed model, closely matched a series of experimental curves obtained at different temperatures. These curves were from an identical set of equations and kinetic parameters for both protium and deuterium. Our numerical kinetic model has the potential to serve as a valuable simulation tool for various applications, such as the design of high-performance hydrogen storage systems and maintenance strategies against hydrogen embrittlement of structural materials. • Protium and deuterium absorption experiments in Ti performed. • The initial Ti temperature in vacuum determines high absorption rate and content. • The initial Ti temperatures of 900 (protium) and 980 °C (deuterium) are optimal. • Temperature dependence indicates the existence of tunneling transport process. • A kinetic model of hydrogen isotope transport for numerical simulations presented. [ABSTRACT FROM AUTHOR]

W-based WTaVCr refractory high entropy alloys (RHEA) may be novel and promising candidate materials for plasma facing components in the first wall and diverter in fusion reactors. This alloy has been developed by a powder metallurgy process combining mechanical alloying and spark plasma sintering (SPS). The SPSed samples contained two phases, in which the matrix is RHEA with a body-centered cubic structure, while the oxide phase was most likely Ta2VO6 through a combined analysis of X-ray diffraction (XRD), energy-dispersive spectroscopy (EDS), and selected area electron diffraction (SAED). The higher oxygen affinity of Ta and V may explain the preferential formation of their oxide phases based on thermodynamic calculations. Electron backscatter diffraction (EBSD) revealed an average grain size of 6.2 μm. WTaVCr RHEA showed a peak compressive strength of 2997 MPa at room temperature and much higher micro- and nano-hardness than W and other W-based RHEAs in the literature. Their high Rockwell hardness can be retained to at least 1000°C. [ABSTRACT FROM AUTHOR]

Fusion energy investigation has stepped to a new stage adopting deuterium and tritium as fuels from the previous stage concentrating hydrogen plasma physics. Special radiation safety issues would be introduced during this stage. In addition to industrial and military uses, tungsten is also regarded as the most promising plasma facing material for fusion reactors. During the operation of fusion reactors, tungsten-based plasma facing materials can be activated via neutron nuclear reaction. Meanwhile, activated tungsten dust can be produced when high-energy plasma interacts with the tungsten-based plasma facing materials, namely plasma wall interaction. Activated tungsten dust would be an emerging environmental pollutant with radiation toxicity containing various radionuclides in addition to the chemical toxicity of tungsten itself. Nonetheless, the historical underestimation of its environmental availability has led to limited research on tungsten compared to other environmental contaminants. This paper presents the first systematic review on the safety issue of emerging activated tungsten dust, encompassing source terms, environmental behaviors, and health effects. The key contents are as follows: 1) to detail the source terms of activated tungsten dust from aspects of tungsten basic properties, generation mechanism, physical morphology and chemical component, radioactivity, as well as potential release pathways, 2) to illustrate the environmental behaviors from aspects of atmospheric dispersion and deposition, transformation and migration in soil, as well as plant absorption and distribution, 3) to identify the toxicity and health effects from aspects of toxicity to plants, distribution in human body, as well as health effects by radiation and chemical toxicity, 4) based on the research progress, research and development issues needed are also pointed out to better knowledge of safety issue of activated tungsten dust, which would be beneficial to the area of fusion energy and ecological impact caused by the routine tungsten related industrial and military applications.

Fusion power plants can meet the energy demands of the world. The high thermal performance of a power generation system is still a challenge and one of the developing trends with a fusion reactor due to the high outlet temperature. A combined cycle system has been optimized to work at high outlet temperature and pressure of fusion power reactor with thermal power of 2500 MWth and its input parameter has been optimized along with impact of partial load to achieve high thermal performance very first time. The combined cycle based on the concept of two stages of expansion in the closed-Brayton-cycle and two reheaters in the Rankine cycle named Schematic-IV with the higher thermal performance of 58.19% as compared to Schematic-I, II and III has been proposed for a fusion power reactor. The increase in more than one expansion stage and reheat stage is not economical because it increased the thermal performance by about half of performance as compared to one. The effect of isentropic efficiency of compressor and heat rate have been investigated to validate the thermodynamic model and calculations. The optimized thermal performance of the combined cycle is 58.19% at a feed water mass flow rate of 592 kgs−1 and steam outlet temperature of 277 °C in the combined cycle. The heat rate decreased and thermal performance increased with the mass flow rate verifying that the thermodynamic model is correct. The thermodynamic analysis of the combined cycle-based nuclear reactor provides insights for better system thermal performance and reveals the effect of key parameters on the system performance.

The modified N50 (N50H) cable-in-conduit conductor CICC jacket developed by China will be applied in China Fusion Engineering Test Reactor (CFETR). Building upon the N50H material, China has developed a corresponding welding material and conducted R&D work on N50H jacket welding. We simulated the N50H jacket welded joint to replicate the entire process of CICC preparation, which includes extrusion, bending, straightening and aging. The magnetic properties of welded joints were evaluated at various temperatures using a physical property measurement system. In addition, the microstructure of the welded joint was studied by various microscopic analysis techniques, and the welded joint showed a stable austenitic structure. Furthermore, mechanical properties of the welded joints were investigated at different temperatures with particular emphasis on their behavior at 4.2 K. The circle-in-square ReBCO conductor jackets welding joint showed a yield strength greater than 1500 MPa, a fracture toughness KIC better than 260 MPa m1/2 and the elongation at break is more than 30 % at 4.2 K. However, a decrease in strength, ductility, and fracture toughness is observed within the welded joints of Nb3Sn conductor jackets after undergoing an aging process. This study will present experimental data on welding joints and discuss the feasibility of N50H as a high-magnetic field jacket material for next-generation fusion reactors.

In this work, hardness and micro- and nano-structure of W–10Cr–0.5Y alloy, which is a promising material for fusion reactors, before and after irradiation with Fe ions with an energy of 5.6 MeV at 500°C were studied. Nanoindentation for hardness measurement and transmission electron microscopy and atom probe tomography for structural changes were used. A formation of Cr clusters with the concentration of Cr in clusters of 52 ± 2 and 77 ± 3 at % for radiation doses of 1 and 10 displacement per atom (dpa), respectively, was detected. An increase in the hardness with increasing the irradiation dose from 1 to 10 dpa was found, which correlates with the increase in the Cr concentration in clusters. [ABSTRACT FROM AUTHOR]

The tritium fuel cycle (FC) is one of the main systems of a fusion plant. Currently, active work is under way in Russia and worldwide to improve FC technologies in order to increase its efficiency and safety. Another area of work is demonstration and testing of technologies for large-scale systems of future fusion plants, which are characterized by large fuel flows and significant amounts of tritium in the systems. In order to plan work on the design and construction of fusion plants, it is necessary to analyze existing technologies and their readiness for use in FC. This article continues the analysis of the readiness of FC technologies of the DEMO-FNS facility in Russia started by the authors earlier. The Technology Readiness Level (TRL) methodology is used for the analysis, according to which technologies in the target application area correspond to different readiness levels from TRL 1 (basic principles observed and reported) to TRL 9 (actual system "flight proven" through successful mission operations). The article analyzes the technologies of tokamak pumping, tritium safety, and creation of materials for tritium handling. It is shown that in Russia vacuum technologies are poorly developed (TRL 2) and critically depend on the use of imported equipment. Methods for the analysis of tritium and its compounds, including technically complex ones, are actively used in Russia at various enterprises, but serial production of the necessary equipment exists for a small list of relatively simple methods, other methods are represented only by prototypes (often single samples), and the application of many analytical methods critically depends on the supply of foreign high-tech equipment. The technologies for creating multilevel protection and for capturing and processing of tritium waste are largely developed (TRL 6 and TRL 5), are applied in tritium laboratories and production facilities, and generally correspond to the world level. The level of readiness of materials for tritium handling (TRL 7) makes it possible to utilize existing technologies to produce equipment, components, and piping for FC systems outside the reactor vacuum vessel while operating at elevated temperatures. In the case of in-vessel components operating under severe neutron irradiation conditions, structural and functional materials should be appropriately tested (TRL 2–3). The level of readiness of the listed technologies is insufficient for application in the FC of the DEMO-FNS facility. It is necessary to increase the level of technology readiness within research and development programs, to create specialized stands for testing and demonstration of technologies, and to create experimental fusion plants for testing and integration of technologies. [ABSTRACT FROM AUTHOR]

Justification of fusion reactor safety is impossible without multiphysics codes that make it possible to model many coupled physical and chemical phenomena. Currently, multimodule integral codes for modeling fusion reactors are actively developed abroad, but there are no such codes in Russia. In this article, the requirements for the EUCLID-F integral code, its modular structure, and the list of modeled phenomena are formulated. The code is developed at the Nuclear Safety Institute of the Russian Academy of Sciences in order to analyze accidents at fusion reactors with magnetic plasma confinement. The design features, hazards, and emergency modes of fusion reactors are briefly described, and a comparison of the EUCLID-F code with foreign counterparts is given. In addition, the requirements for the unified database and high-fidelity codes necessary for development and performance of calculations by the integral code are formulated. Taking into account the described requirements when developing the integral code will make it possible to use it for analyzing the majority of design and beyond design basis accident scenarios. The code developed in accordance with the above requirements will significantly surpass foreign counterparts in terms of the number and variety of calculation modules and simulated phenomena. The availability of this code in Russia will significantly contribute to the development of fusion technologies. [ABSTRACT FROM AUTHOR]

Photoluminescence (PL) properties of Er2O3 specimens were examined by using visible lasers (532 nm and 635 nm) and a UV LED light source (365 nm) to investigate the applicability for irradiation damage monitoring of materials in fusion reactors. Both in the laser induced and UV light induced PL spectra, green (510–590 nm) and red (630–725 nm) luminescence was observed. In the spectrum measurements on specimens with different crystallinities, it was confirmed that an intensity of the red luminescence weakened significantly compared with that of the green luminescence in an Er2O3 specimen with a lower crystallinity. The results indicate that the PL measurements of Er2O3 materials could be applicable for the irradiation damage monitoring in fusion reactors. The luminescence property of ion beam irradiated Er2O3 showed that information of irradiation damages could be kept up to ∼ 300 ℃ and almost recovered at 700 ℃. Based on the obtained luminescence properties, positions in a fusion reactor where Er2O3 materials could be used as irradiation damage sensors are proposed. Changes in PL spectra at high temperatures up to ∼ 400 ℃ indicate the possibility that the Er2O3 materials might be applicable also for temperature monitoring of in-vessel components during reactor maintenance periods.

Beryllium-based intermetallic compounds, such as Be12Ti, are increasingly being considered as a material capable of replacing pure beryllium in structural elements of fusion reactors. Be12Ti is considered as a neutron breeder material, a structural part of the Helium Cooled Pebble Bed of the DEMO reactor. It is expected that the replacement of beryllium by Be12Ti will make it possible to reduce the capture of tritium in the blanket without a significant decrease in the neutronic characteristics. Unlike beryllium, beryllides have relatively recently begun to be considered for use in nuclear and thermonuclear facilities, so the radiation resistance of these compounds remains little studied. This paper presents the experimental results on effect of low temperature neutron irradiation to properties of titanium beryllide samples manufactured by industrial technology in the Ulba Metallurgical Plant (UMP, Kazakhstan). The manufactured samples before and after irradiation were analyzed by scanning electron microscopy (SEM), X-Ray diffraction (XRD), hydrostatic weighing method, dimension method and microhardness measurement by Vickers method.

The deuterium-tritium fusion reaction is the fastest commercially achievable artificially controlled fusion reaction. However, tritium, as a fuel and neutron radiation-activated material, poses radioactive safety challenges. To simulate the migration of tritium and radioactive dust in the environment under accident conditions, a dispersion simulation program called ACCTRI (ACCidental model for TRItium release) has been developed based on the modified Gaussian multi-puff model, considering dry and wet deposition and tritium re-emission effects. The concentration and dose results calculated using ACCTRI are very close to those calculated using UFOTRI and HotSpot, with a maximum difference of no more than one order of magnitude. A comparison with the experimental results reveals that the results of ACCTRI with tritium buoyancy correction are close to the experimental data. ACCTRI has good accuracy and can be used as a reference for environmental radioactivity safety in fusion reactor site selection and hypothetical accident analysis.

BackgroundWhen fusion nuclear reactor is in operation, neutron activation causes a large number of radioisotopes. Activation calculation is a very important step in both of reactor shielding calculation and radiation safety analysis.PurposeThis study aims to develop the capability of transport-activation coupling calculation for fusion reactor based on the Monte Carlo code cosRMC.MethodsFirstly, the built-in burnup solver "Depth" of cosRMC was employed to develop transport-activation internal coupling calculation function under fixed source mode with embedded calculation of activation related nuclide single group reaction cross-sections in neutron transport process without the transmission of neutron spectra to external activation programs. Then, the developed code was applied to the activation calculations of the first wall (FW) material steel and plasma facing component (PFC) material tungsten of the Chinese Fusion Engineering Testing Reactor (CFETR) using continuous energy cross-sections and multi group cross-sections, respectively, and calculated results were compared and verified with that of the activation program ALARA.ResultsThe comparison results of activation calculation of FW steel and PFC tungsten in CFETR show the consistency between cosRMC-based internal coupling method and ALARA program, which preliminarily verifies the correctness of the transport activation internal coupling calculation function of the developed cosRMC program.ConclusionThe developed cosRMC-based internal coupling method can dynamically update neutron spectra and material information, and use continuous energy cross-sections for reaction rate calculation, obtaining reaction cross-sections related to the geometry and energy spectra of actual problems, thus accurately considering the influence of resonance zone nuclear cross-sections.

BackgroundThe amount of gaseous tritium stored and released in fusion reactor is higher than that in current fission reactors, so that tritium is an important source of potential radioactivity in fusion reactor.PurposeThis study aims to investigate the environmental impact of gas tritium emission from fusion reactor for achieving the safety and environmental friendliness of fusion reactor in the future.MethodsTypical factory sites along the eastern coast of China with the highest frequency east wind direction was selected as the research object, and the Gauss model was employed to predict the atmospheric dispersion of gaseous tritium release and the dry deposition of tritium gas (HT), soil oxidation and re-evaporation of HTO. The radiation dose of 1 g HT in the case of short-term released from fusion reactor to the public in the surrounding environment was calculated.ResultsCalculation results show that the effective dose of inhalation internal irradiation of HT released at 10 m height for adults at 500~3 000 m west of the release point ranges from 0.38 mSv to 0.1 mSv. The dose caused by the re-evaporation effect of HTO at different distances is the main source of the dose of gaseous tritium. The proportion of the HT deposited to soil being oxidized to HTO and the atmosphere condition are the key parameter determining the effective dose of the tritium gas.ConclusionsThe study shows that the effective dose of HT released from fusion reactor to public is higher than which released from fission reactor, hence further attention to the environmental impact of the tritium is needed in the research on the fusion reactor subsequently.

This paper proposes a design method for a miniature air–land amphibious inspection drone (AAID) to be used in the latest compact fusion reactor discharge gap observation mission. Utilizing the amphibious function, the AAID realizes the function of crawling transportation in the narrow maintenance channel and flying observation inside the fusion reactor. To realize miniaturization, the mobile platform adopts the bionic cockroach wheel-legged system to improve the obstacle-crossing ability. The flight platform adopts an integrated rotor structure with frame and control to reduce the overall weight of the AAID. Based on the AAID dynamic model and the optimal control method, the control strategies under flight mode, hover mode and fly–crawl transition are designed, respectively. Finally, the prototype of the AAID is established, and the crawling, hovering, and fly–crawling transition control experiments are carried out, respectively. The test results show that the maximum crawling inclination of the AAID is more than 20°. The roll angle, pitch angle, and yaw angle deviation of the AAID during hovering are all less than 2°. The landing success rate of the AAID during the fly–crawl transition phase also exceeded 77%, proving the effectiveness of the structural design and dynamic control strategy.

During reactor operation, the divertor must withstand unprecedented simultaneous high heat fluxes and high-energy neutron irradiation. The extremely severe service environment of the divertor imposes a huge challenge to the bonding quality of divertor joints, i.e., the joints must withstand thermal, mechanical and neutron loads, as well as cyclic mode of operation. In this paper, potassium-doped tungsten (KW) is selected as the plasma facing material (PFM), oxygen-free copper (OFC) as the interlayer, oxide dispersion strengthened copper (ODS-Cu) alloy as the heat sink material, and reduced activation ferritic/martensitic (RAFM) steel as the structural material. In this study, a vacuum brazing technology is proposed and optimized to bond Cu and ODS-Cu alloy with the silver-free brazing material CuSnTi. The most appropriate brazing parameters are a brazing temperature of 940 °C and a holding time of 15 min. High-quality bonding interfaces have been successfully obtained by vacuum brazing technology, and the average shear strength of the as-obtained KW/Cu and ODS-Cu alloy joints is ∼268 MPa. And a fabrication route for manufacturing the flat-type divertor target based on brazing technology is set. For evaluating the reliability of the fabrication technologies under the reactor relevant condition, the high heat flux test at 20 MW/m2 for the as-manufactured flat-type KW/Cu/ODS-Cu/RAFM mockup is carried out by using the Electron-beam Material testing Scenario (EMS-60) with water cooling. This paper reports the improved vacuum brazing technology to connect Cu to ODS-Cu alloy and summarizes the production route, high heat flux (HHF) test, the pre and post non-destructive examination, and the surface results of the flat-type KW/Cu/ODS-Cu/RAFM mockup after the HHF test. The test results demonstrate that the mockup manufactured according to the fabrication route still have structural and interfacial integrity under cyclic high heat loads.

In fusion reactors, many blanket concepts are designed with water as a coolant to transfer high-density heat from the fusion reaction out of the reactor core. The coolant temperature and pressure are maintained as the validated use in water-cooled fission reactors. However, the flow channel in a water-cooled blanket is independent of each other, and there is no flow mixing between coolant channels. Therefore, flow instability may occur in the independent parallel channels in a water-cooled blanket due to its unique structure and heat distribution, especially under the high heat flux caused by plasma rupture. In this study, the parametric analysis of density wave instability is performed using a thermal-hydraulic code developed for independent parallel channels based on the homogeneous model for the two-phase flow. The parallel-channel system in a water-cooled ceramic breeder (WCCB) blanket of the China Fusion Engineering Experimental Reactor (CFETR) is established for its first wall structure. A small disturbance is introduced into the system to determine if it is stable under different conditions. It is found that the channel number has no obvious influence on the prediction of the flow instability boundary. Therefore, the two-channel system is adopted to investigate the influence of different parameters, such as the pressure, resistance, flow rate, and inclination, on the flow instability boundary of the parallel-channel system in the CFETR WCCB blanket. The results show that flow instability occurs more easily in this study compared to the traditional instability analysis, especially under high-pressure conditions. In general, conditions of high pressure, large flow rate, and no inclination can stabilize the system, while the influence of resistance is quite different under different conditions of resistance and pressure. The research work indicates that more attention should be paid to the joint influence of different parameters for the water-cooled blanket during its design and operation.

Developing cryogenic structural materials with high strength and toughness is a challenge for the high-field superconducting magnet of the future fusion reactor. 0.2% proof stress should be over 1500 MPa and the fracture toughness KIC should be better than 130 MPa·m1/2 at 4.2 K for future cable-in-conduit conductors (CICC) jackets. Thus the cryogenic structural materials of modified 316LN, JK2LB, Incoloy 908, and JJ1 developed by ITER do not meet the requirements. Based on Nitronic 50 (N50) super-austenitic stainless steel material, the modified N50 (N50H) developed in China shows excellent low-temperature mechanical properties by completely eliminating δ ferrite and strictly controlling the carbon content of the alloy. Meanwhile, the jackets of CICC were prepared with N50H material for the first time, and some R&D work has been done. The N50H jackets have been compacted to final dimensions, bent to a radius of r = 2000 mm, straightened, and aging treated. The yield strength of N50H circle-in-square jackets is greater than 1550 MPa, and the fracture toughness KIC is better than 150 MPa·m1/2 at 4.2 K. The fatigue crack growth rate (FCGR) of the N50H jacket is lower than that of the modified 316LN, JK2LB and close to Incoloy 908. This study will present experimental data of modified 316LN, JK2LB, Incoloy 908, JJ1, and N50H. Moreover, this paper discusses the feasibility of N50H as a high-magnetic field cryogenic structural material for next-generation fusion reactors.

Currently, international nuclear fission reactors producing medical isotopes face the problem of shutdown and maintenance, decommissioning, or dismantling, while the production capacity of domestic research reactors for medical radioisotopes is inadequate, and the supply capacity for medical radioisotopes faces major challenges in the future. Fusion reactors are characterized by high neutron energy, high flux density, and the absence of highly radioactive fission fragments. Additionally, compared to fission reactors, the reactivity of the fusion reactor core is not significantly affected by the target material. By building a preliminary model of the China Fusion Engineering Test Reactor (CFETR), a Monte Carlo simulation was performed for particle transport between different target materials at a fusion power of 2 GW. The yields (specific activity) of six medical radioisotopes (14C, 89Sr, 32P, 64Cu, 67Cu, and 99Mo) with various irradiation positions, different target materials, and different irradiation times were studied, and compared with those of other high-flux engineering test reactors (HFETR) and the China Experimental Fast Reactor (CEFR). The results show that this approach not only provides competitive medical isotope yield, but also contributes to the performance of the fusion reactor itself, e.g., tritium self-sustainability and shielding performance. [ABSTRACT FROM AUTHOR]

In-vessel plasma facing components in fusion reactors such as ITER and DEMO experience high thermal loads and require active cooling, for which water is one possible coolant. The solution where plasma facing components utilise tungsten blocks as sacrificial armour has a joint internal structure of cooling tubes made from the copper base (∼99 % Cu) alloy CuCrZr. This paper concerns the testing of CuCrZr tubes with respect to flow-assisted corrosion (FAC) at water velocity in the range 8–10 m/s and temperature in the range 150 °C–250 °C. The FAC rate was evaluated by gravimetry, microscopy, and sampling of the water from the re-circulating test system. When the electrochemical potential of the CuCrZr specimens was kept in the reducing range where Cu is thermodynamically stable as a metal, no FAC could be observed on the specimens. At electrochemical potentials that promotes oxidation of Cu through presence of an oxidizer the FAC rate of a straight CuCrZr tube was estimated to be in the order of 400 µm per full power year.

Due to the expression of “reduced activation”, the severe radioactivity of tantalum (Ta) after neutron irradiation in early decay years is easily neglected. In the presented study, after an irradiation of only 0.3 dpa by fission neutron and followed by a cooling stage of 305 days, the RAFM steel specimen with a weight of 0.8 g possessed drastic radioactivity of 4 mSv/hour at a measuring distance of 20 cm. It was ascertained that the high radioactivity was generated by 182Ta. The severe radioactivity seriously increased the danger of material experiments, even though the original Ta addition of the steel was only 0.13 wt%. Similar deteriorative radioactivity of Ta can be generated in both fission and fusion neutron irradiations. We appeal to seriously deliberate the radioactivity of Ta and control the Ta content in the RAFM steel.

In order to complete the tritium production of irradiated tritium breeder in the solid tritium breeder reactor of fusion reactor, in addition to the conventional ionization chamber, a high-precision gas chromatography on-line detection and analysis method of neon carrier gas was established in this work. The tritium production in the system was verified by measuring the helium production in the tritium production circuit. Thus, a new tritium production rate measurement and verification method was provided for the irradiation tritium production performance of tritium producing cladding solid breeder materials in fusion reactor. It’s necessary to establish the method of analyzing trace 4He, H2 and impurity components in Ne, in order to complete the test of real-time online detection of the system, developing three detectors and five chromatographic columns. The results show that the developed chromatographic analysis system can realize the detection and analysis of 4He, H2 and impurity components in high-purity Ne, the detection limits of H2 and 4He can reach 1.0×10-6 and 5.9×10-6 respectively, the relative standard deviation(sr) of each component content and peak area are less than 5.0%(n=6), and the linear correlation coefficient r2 is greater than 0.99, indicating that the detection method has good repeatability. According to the on-line test of multicomponent gas in Ne, the measurement repeatability in single period and multi period is good, which can provide an analytical means for the verification of tritium production rate in the irradiation tritium production assessment system, and then provide technical support for the formal entry into the reactor to obtain irradiation data and tritium balance.

Probabilistic Safety Assessment (PSA) is an important method to evaluate the safety concerns of Nuclear Power Plants (NPPs) from probability perspective. The Reliability, Availability, Maintainability, Inspectability (RAMI) approach was promoted by the International Thermonuclear Experimental Reactor (ITER) organization to evaluate system reliability and availability, and reduce the technical risks to an acceptable level from probability perspective for fusion devices. Both PSA and RAMI adopt probabilistic approach to evaluate the failure issues of nuclear systems. In this contribution, the analysis purposes, methods, objects, and quantitative analysis indexes of PSA and RAMI were firstly compared. Then, an integrated safety assessment method based on PSA and RAMI for fusion reactors was proposed, and the case study of in-vessel LOCA accident for fusion reactors was performed to demonstrate the effectiveness of the proposed method. The application results indicate that the proposed method not only can perform PSA, but also RAMI analysis for fusion reactors. The proposed method integrates the characteristics of PSA and RAMI, breaks the barrier between PSA and RAMI, preliminarily solves some problems of PSA for fusion reactors, and enriches the connotation of PSA for fusion reactors.

The electrical conductivity of the α-Al2O3 layer formed on FeCrAl alloy APMT (Fe-21Cr-5Al-3Mo) in air atmosphere at 1273 K and 1373 K was measured in a temperature range between room temperature and 1073 K. Low electrical conductivity of the α-Al2O3 layer was clarified in the measurement temperature range. Low electrical conductivity of the ZrO2 layer formed on the Zr metal in air atmosphere at 873 K was also measured. Numerical simulation was performed with three-dimensional thermo-fluid code to clarify the reduction of magnetohydrodynamic (MHD) pressure drop with the electrical insulating α-Al2O3 and ZrO2 layers in liquid blankets of magnetic confinement fusion reactors. The MHD pressure drop is significantly reduced when the four inner surfaces of the duct are electrically insulated by the oxide layers. The electrical insulation of the three inner surfaces is also effective to reduce the MHD pressure drop when the magnetic field is parallel to the conductive wall of the flow duct. However, the MHD pressure drop is induced in the three-surface insulation duct when the magnetic field makes an angle with the conductive wall of the duct.

The paper discusses the prospects of an alternative way to close the fuel cycle, without recycling highly radioactive spent fuel. The essence is the potential production of the fissile isotope 233U from the isotope 232Th in a hybrid fusion reactor (HFR). The feasibility of using a hybrid fusion plant as a fuel accumulator in the nuclear fuel cycle, and the potential need of the system for hybrid fusion reactors necessary for its closing are estimated. A system in which two technologies, fission and fusion, working together, avoid the difficulties that arise as a result of the independent implementation of each of the technologies under consideration is simulated. [ABSTRACT FROM AUTHOR]

This article describes the development of the DEMO-FNS hybrid (fusion–fission) reactor with DT fusion capacity of 40 MW in Russia. Operation of the reactor requires the development of systems of the fusion fuel cycle (FC). They are based on technologies of tritium and deuterium handling, which are being developed and applied in various fields of science and engineering. The necessity to improve the tritium technologies in Russia is dictated by conversion to radically higher gas flows and tritium reserves in the fuel cycle of hybrid and fusion systems under conditions of limited import of dual purpose technologies. In order to select reliable technologies and intensify developments in critical issues of FC, it is required to analyze the technology readiness level. We have estimated the readiness of existing technologies of tritium and deuterium handling in Russia for use in the fuel cycle of DEMO-FNS. The analysis is based on the Technology Readiness Level (TRL) method, according to which each technology is assigned a readiness level from TRL 1 (basic technology principles have been demonstrated) to TRL 9 (technology has been verified by successful operation). The technologies of membrane separation of hydrogen-containing gas mixtures, cryogenic hydrogen rectification, chromatographic separation of hydrogen isotopes, cryogenic adsorption separation, gas detritiation in scrubber, and CECE process (Combined Electrolysis and Catalytic Exchange) have been discussed. Other FC technologies will be considered in our further publications. The listed technologies have been verified in Russia and are used in various fields of industry and science. However, operational conditions of the technologies differ from the planned parameters of the DEMO-FNS fuel cycle, for which most technologies are at the stage of development (TRL 4–6); some technologies, such as cryogenic adsorption separation and chromatographic processes, comply with the research stage (TRL 1–3). The state of these technologies is "below" or "complies with the world level." Further development of the considered technologies requires specialized facilities and test benches, which would make it possible to optimize their combined use under conditions simulating operation of a fusion facility. [ABSTRACT FROM AUTHOR]

A list of requirements for the measurements of characteristics of high-temperature plasma that ensure safe operation and protection of the tokamak with reactor technologies, along with control of the plasma column and investigation into fusion-plasma physics, is formulated. Special attention is paid to measurements the results of which are needed for the development of plasma technologies of future quasi-stationary fusion reactors. Plasma parameter ranges, including the requirement for the resolution with which these parameters must be measurement, are defined. These requirements represent the initial data for choosing the systems of measurements, development of construction, and subsequent integration of diagnostics in the system. Classification of diagnostics depending on the problems to be solved and the stages of diagnostic complex development of the tokamak with reactor technologies is proposed. [ABSTRACT FROM AUTHOR]

The operation of a fusion reactor on a fuel alternative to the D–T (D–D, Cat–DD, D–3He, p–11B, and p–6Li) requires a higher fuel mixture temperature and a longer energy confinement time in the plasma. It is shown that for a fixed fusion reaction power, an increase in the plasma parameters peaking reduces the power required for additional plasma heating. In addition, the peaking of the plasma parameters reduces the radiation loss in the plasma. All this softens the requirements to the operation conditions of fusion reactors. [ABSTRACT FROM AUTHOR]

The hydrogen infrastructure involves hydrogen production, storage and delivery for utilization with clean energy applications. Hydrogen ingress into structural materials can be detrimental due to corrosion and embrittlement. To enable safe operation in applications that need protection from hydrogen isotopes, this review article summarizes most recent advances in materials design and performance characterization of barrier coatings to prevent hydrogen isotopes' absorption ingress and permeation. Barriers are crucial to prevent hydride formation and unwanted hydrogen effects to increase safety, materials' lifetime and reduce cost for applications within nuclear and renewable energy. The coating may be applied on a material that requires protection from hydrogen pick-up, transport and hydride formation in hydrogen storage containers, in pipelines, spent nuclear fuel storage or in nuclear reactors. While existing, commercial coatings that have been much in use may be satisfactory for various applications, it is desirable to evaluate whether alternative coating concepts can provide a greater resistance to hydrogen isotope permeation along with other improved properties, such as mechanical strength and thermal resistance. The information presented here is focusing on recent findings within the past 5–7 years of promising hydrogen barriers including oxides, nitrides, carbon, carbide, MAX-phases and metals and their mechanical strength, hydrogen pick-up, radiation resistance and coating manufacturing techniques. A brief introduction to hydrogen permeation is provided. Knowledge gaps were identified to provide guidance for material's research prospects. [ABSTRACT FROM AUTHOR]

Chinese Fusion Engineering Test Reactor (CFETR) has been proposed to bridge the technology gap between the International Thermonuclear Experimental Reactor (ITER) and the Fusion Demonstration Reactor (DEMO). As the most crucial nuclear component, fusion blanket undertakes the functionalities of tritium breeding, energy extraction and radiation shielding. The design of fusion blanket directly determines whether the fusion reactor can operate in a safety and steady way. In the design of fusion blanket, the neutronics and thermal-hydraulics are the most associated aspects. Actually, the neutronics and thermal-hydraulics are closely coupled to each other in the fission reactor. The effect caused by the interaction between neutronics and thermal-hydraulics is called the neutronics/thermal-hydraulics coupling effect. In the fission reactor, the neutronics/thermal-hydraulics coupling effect will lead to significant impact on both of neutronics and thermal-hydraulics, e.g., change the effective multiplication factor, and influence power distribution and temperature distribution. Therefore, the neutronics/thermal-hydraulics coupling effect must be taken into account in the design work of fission reactor. However, the influences of the neutronics/thermal-hydraulics coupling effect are still not clear in the fusion reactor. The key point to perform the neutronics/thermal-hydraulics coupling calculation of fusion blanket is to choose rational solvers and coupling method. Considering the complicated structure of the fusion blanket, Monte-Carlo code MCNP and computational fluid dynamics code FLUENT, which have good geometric adaptability, are selected as the neutronics solver and the thermal-hydraulics solver, respectively. Nevertheless, the direct three-dimensional coupling calculation is still difficult because of the complicated geometric mapping relationship and huge amount of calculation. Thus, a hybrid 3D-1D-2D coupling method is used to deal with the spatial mapping between neutronics model and thermal-hydraulics model. Moreover, the pseudo material method is adopted to efficiently handle the variation of neutron cross sections under different temperatures. In this paper, the main conceptual blanket designs of CFETR, including the helium cooled solid blanket design under 200 MW, and the water cooled solid blanket designs under 200 MW and 1.5 GW, were selected as the research objects to study the neutronics/thermal-hydraulics coupling effect on the CFETR. Above all, the neutronics sensitivity analysis of material temperature of fusion blanket was conducted to demonstrate how temperature influenced the tritium breeding capability and nuclear heat deposition. The results of sensitivity analysis indicate that the thermal scattering effect of beryllium and the density of water are the decisive factors of temperature effect. Subsequently, the neutronics/thermal-hydraulics coupling calculation was carried out. The coupling results show that, in the helium cooled solid blanket, the neutronics/thermal-hydraulics coupling effect is small and can be ignored, and in the water cooled solid blanket, the neutronics/thermal-hydraulics coupling effect will slightly influence the tritium breeding capability and temperature distribution.

The research presents an experiment with a model of an electronic generator for energy injection into the plasma of a fusion reactor. By recording a non-ionizing field in the vicinity of a fusion reactor, it was determined that this field has an extremely high growth rate. At the site of the maximum intensity of the field of non-ionizing radiation, commercial surge arresters with a flexible model of surge arresters were used for experimentation. It has been found that the commercial surge arresters have an efficiency of about 20%. For the efficiency of the flexible model, it was found to be slightly less than 40% (and to be achieved by the application of alpha particle radiation). Since neither of these efficiencies guarantee reliable operation of the gas-filled surge arrester, it was concluded that essential electronics in the vicinity of the fusion generator must be protected. However, since this protection can only be implemented in a fusion reactor, the fact remains that the environment of such a reactor is extremely contaminated with non-ionizing radiation. Commercial surge voltages are isolated for testing since the protection of electronic circuits from fast overvoltages is a critical point for the functioning of modern electronics.

• Proposed a design method for integrated laser cutting/welding robots in pipes. • Proposed a hierarchical control system architecture for cutting and welding robots. • Manufactured cutting and welding robot test prototype systems. • The experiments verified the correct design and control of the cutting and welding robot. Based on the specifications and layout of piping systems related to CFETR design, this paper studies automatic in-pipe laser cutting and welding robots (CWR) in confined spaces. Firstly, a sizeable slender ratio laser system design principle is proposed according to the demand for pipe cutting/welding inside the vacuum chamber. Based on the YLS4000 laser system, the design of an integrated laser cutting/welding head suitable for narrow spaces is completed. Then, the design of the whole CWR machine is achieved by matching the coordinated moving platform, control system, fixture, and other equipment. The fixture alignment operation and cutting/welding workflow are analysed for the CWR work process. A hierarchical control system architecture is proposed for motion control requirements, and the relationship between the human-computer interaction layer, the control layer and the driver layer is clarified. Finally, the CWR test prototype is produced, and the cutting and welding experiments of flat plate and round pipe are carried out, respectively. The results show that CWR has excellent cutting and welding results, and the pipe welding leakage rates are all less than 8.5 × 10 − 12 pa · m 3 / s , which proves the usability of CWR in the future maintenance of vacuum chamber pipes. [ABSTRACT FROM AUTHOR]

• This paper focuses on the defect of high driving power caused by the low density of helium coolant in the tritium breeding blanket of CFETR. Based on extensive research, a novel approach is proposed to mix supercritical CO2 into the blanket to optimize the driving power and improve the energy conversion efficiency. The innovations are summarized as follows. • The change of coolant property as a function of temperature and the constant property parameters show that the change of property with temperature cannot be ignored.Therefore, the physical parameters of coolants with different mixing ratios are fitted, and a UDF program is developed to calculate them. • The necessary fan driving power for different gas mixtures is calculated based on the calculation results of the first wall under different coolants and the data of Helium coolant in the whole blanket, and the energy proportion of the fan driving power in the total thermal power of the reactor is determined. • Combined with the energy conversion system of the secondary loop, the thermal-electric conversion efficiency of the whole reactor is calculated. It is found that the mixture with the lowest fan driving power may not necessarily have the highest thermal-electric conversion efficiency, which provides some value for further research. The first wall of the tritium breeding blanket is an important high-heat flux component in fusion reactors, and its thermal analysis is critical. Helium is chosen as the coolant for the CFETR tritium breeding blanket due to its good compatibility with structural materials and excellent thermal properties. However, its low density requires a significant amount of drive power. After extensive feasibility research, using He-CO 2 mixture as coolant is one of the main optimization directions. In this paper, a new method is proposed to consider the variation of thermal properties parameters of the He-CO 2 mixture with temperature. A fitting formula for the He-CO 2 mixture at different ratios under 12MPa is provided, and the mixture is simulated in the commercial CFD software Fluent by compiling a UDF program. The results show that the optimum He mole fraction in the He-CO 2 mixture is 60%, which reduces the required drive power by 89.36% compared with using pure He as coolant and by 81.57% compared with using CO 2.With further programming calculations and considering the effect of the secondary circuit, it is calculated that the highest energy conversion efficiency is 29.43%.This study provides valuable insights into the thermal-hydraulic performance of the tritium breeding blanket with a binary gas mixture coolant in fusion reactors. [ABSTRACT FROM AUTHOR]

• Propose a design scheme for the vacuum chamber inspection system with air-ground separation. • Designing safety cables for drone emergency rescue and precision takeoffs and landings. • Proposing a 3D lidar-based real-time localization technique for dark environments. • Flight validation is conducted inside the PF6 superconducting coil shells. This paper proposes an air-ground separation drone endoscopic system (DES) design scheme for the fusion reactor internal inspection task. A foldable rotor drone with vertical takeoff and landing capability is used as the main body of DES. The nest platform is designed separately from the drone to reduce the weight. Meanwhile, the nest platform can be installed with a safety cable and 2D code to realize emergency rescue and precise take-off and landing. DES adds 3D lidar sensors for 360° stereo sensing, realizing real-time positioning in dark or even lightless environments in vacuum chambers. The first-generation DES prototype is developed, and the flight test is performed inside the PF6 superconducting coil shell by fusing 3D lidar point cloud data with visual camera image data. The results show that the position error of DES is stabilized within ±0.05 m during the hovering phase. The error is mainly due to the narrow space inside the PF6 coil shell, which produces a perturbed flow field on the DES. In general, the DES accomplished circumferential autonomous flight control and 3D point cloud data acquisition in a confined space, proving the feasibility of drone inspection in vacuum chambers. [ABSTRACT FROM AUTHOR]

Based on the design requirements of the first stage of the China Fusion Engineering Test Reactor (CFETR), the activation performance calculation and analysis of the components in the dual function lead-lithium blanket were performed using a neutron transport design and safety evaluation software system (SuperMC) with the fusion evaluation database JEFF3.2. A coupled burnup and transportation calculation method was used to calculate the radioactivity, residual decay heat, dose rate, and potential biological hazards of each component in the equatorial surface of the fusion reactor at different times after shutdown. The waste disposal problem of the tritium breeding blanket after the decommissioning of the fusion reactor was analyzed according to the relevant nuclear waste disposal standards in the safety and environment assessment of fusion power strategy. Results of the analysis show that under the condition of normal operation of 200 MW for 10 years, all components of the breeding blanket in the CFETR can satisfy simple recycling standards and meet the requirements of the first stage of CFETR radioactive waste treatment after 50 a of cooling.

The possibilities are considered of using the Thomson scattering diagnostics of core and edge plasmas in the tokamak with reactor technologies, which is under design. The problems are described that can be solved using the Thomson scattering diagnostics, including the possibility of controlling the plasma current profile. Technical requirements for the diagnostics are formulated. The possibilities are analyzed of its arrangement in the tokamak vacuum chamber. The accuracies are estimated of measuring the electron temperature and density of the plasma created in the tokamak. Particular attention is paid to ensuring the operability of the proposed diagnostics in the reactor regime of the tokamak operation. [ABSTRACT FROM AUTHOR]

Reasonable and feasible reliability index allocation is significant for improving safety and reducing costs for fusion reactors. Although the reliability index is allocated when designing some key systems, reliability allocation is not implemented from the overall layer to the system and component layers for fusion reactors. In this contribution, fusion reactors are divided into different layers by applying the Analytic Hierarchy Process (AHP); these layers include the overall layer, system layer and component layer, failure mode layer, etc. Combining the Probabilistic Safety Assessment (PSA) model for fusion reactors and AHP, the reliability index is allocated from the overall layer to the system and component layer. Firstly, a new reliability index allocation method based on PSA and AHP for fusion reactors is proposed. Secondly, the PSA model for an in-vessel LOCA accident of fusion reactors is selected as a case study to demonstrate the applications of the proposed method. Lastly, the relationship between PSA, items safety classification and reliability allocation are discussed. The allocation results indicate that the proposed method provides a more systematic reliability allocation scheme and contributes to improving the safety and economy of fusion reactors. [ABSTRACT FROM AUTHOR]

Experiments concerning the effect of helium (He) plasma exposure on deuterium (D) plasma-driven permeation through tungsten (W) foils in a linear plasma facility have been performed. 0.05 mm thick W foils were exposed to ∼2 Ă— 1020 mâˆ'2 sâˆ'1 He plasma with various fluences at 883 K. After He irradiating, D permeation tests were performed for the samples and retention was also measured by high-resolution thermal desorption spectroscopy. It was observed that He pre-irradiation resulted in a significant reduction of D permeation and retention in W. Microstructure observation indicated that the surfaces of the samples after He irradiation turned rough and He nanobubbles were formed near the surface. Defective structures including He nanobubbles likely enhance D reemission and accordingly reduce the permeation and retention in He pre-irradiated W. [ABSTRACT FROM AUTHOR]

The brazeability of four different alloys (Au, Cu, and two Ag-based alloys) was evaluated for their use as filler materials in joints between EUROFER 97 and tungsten for its application in future fusion reactors. The study aims to analyze the operational brazeability in terms of deep microstructural analysis and mechanical behavior. In general, high metallic continuity was observed for all filler compositions. In the case of the joints brazed with the Au-based filler alloy, a homogeneous microstructure based on an Au-Pd-Fe-Ni solid solution is obtained. The use of Ag-based filler alloys produced a solid solution phase at the EUROFER97-braze interface, and a Ag-based phase in contact with the tungsten base material. Finally, with the cupronickel filler alloy, a braze constituted by two different Cu-Ni-Fe solid solution phases is obtained. Regarding the mechanical characterization, the Cu-based filler shows a lower hardness value, while the higher values were obtained with one of the Ag-based filler alloy. In the case of the shear tests, a maximum 304 ± 57 MPa strength is obtained for Au-based filler alloy brazed at 1171 ºC due to the combination of a homogeneous and toughness microstructure and the lack of intermetallic compounds in the braze. [Display omitted] [ABSTRACT FROM AUTHOR]

A promising magnetic fusion steady state reactor is Chinese Fusion Engineering Testing Reactor (CFETR) and a different power conversion system is required than tokamak reactor. The control and utilization of high outlet temperature and thermal power generated by the fusion reactor efficiently are the main challenges in the field of thermo-fusion technology. An efficient and optimized power conversion system is an urge for the thermo-fusion energy system with a candidate working fluid under high temperature and pressure. Five cycle topologies based on the Brayton cycle have been designed such a s multi stage expansion and recompression cycle with intercooling, recompression cycle with intercooling, partial expansion with multistage compression, partial expansion and partial recompression with cooling across range of turbine and compressor inlet temperatures. Brayton cycle with different system configurations have been simulated with He-gas and supercritical Carbon Dioxide (CO 2) to conceive high outlet temperature of CFETR about 500 °C and thermal power of 200 MW th for better thermal performance by using REFPROP, EES and analytical model has been developed with MATLAB. The supercritical CO 2 has better thermal performance about 36.18 % as compared to He-gas has 29.81 % under same thermal conditions. The Schematic-I with addition of preheat and recompression enhance the thermal performance about 2 % and has been proposed for CFETR. The effect of change in pressure of the system, effect of the working fluid, isentropic efficiency, heat rate and back to work ratio on the thermal performance and network output have been analyzed and optimized based on analytical model with the MATLAB. The system net power is increased from 69.20 MW to 79.71 MW, heat rejected by the system is decreased from 134.46 MW to 127.06 MW and thermal performance is increased about 34.62 %–39.85 % with pressure from 24 MPa to 28 MPa. Using multi stage expansion and recompression cycle with intercooling instead of other topologies can improve the thermal performance up to 6.1 % depends upon the operation conditions and working fluid. The heat rate of the system decreased while back work ratio increased with the pressure and it reveal that calculations are correct. • The steady state Chinese Fusion Engineering Testing Reactor (CFETR) is a promising candidate to meet world energy demands. • The control and utilization of high outlet temperature and thermal power are main challenges in field of thermo-fusion technology. • SCO 2 Brayton cycle with multistage expansion, compression and intercooling has been designed and proposed for CFETR. [ABSTRACT FROM AUTHOR]

• LiF-rich Flinabe showed good heat transfer metrics, though with high melting points. • Flinabe shows heat transfer characteristics comparable to conventional LiF–BeF 2. • LiF–NaF–BeF 2 = 33–29–38 mol% is the optimal composition for heat transfer performance. In this study, we evaluated the dependence of the heat transfer properties of a molten salt mixture (LiF–NaF–BeF 2 —Flinabe) on its composition ratio to determine its applicability as a coolant in a nuclear system. Specifically, we evaluated the density, specific heat, viscosity, and thermal conductivity of Flinabe using molecular dynamics simulation and calculated the Prandtl number and figure of merit heat transfer metrics. The calculated density, specific heat, and viscosity differed from the experimental values by approximately 20 %. Although we could not directly compare the calculated thermal conductivity of Flinabe due to a lack of accurate experimental measurements, we discovered a temperature dependence consistent with that of other salts. Overall, the findings of the present study revealed that light LiF-rich molten salts have favorable heat transfer characteristics. The ternary system LiF–NaF–BeF 2 = 33–29–38 mol% has a relatively low Prandtl number and figure of merit compared to the binary system LiF–BeF 2 , and the melting point is 25 K lower than that of a binary salt. We concluded that LiF–NaF–BeF 2 = 33–29–38 mol% is the optimal composition for a heat transfer medium for nuclear power systems. [ABSTRACT FROM AUTHOR]