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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]

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]

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]

• 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]

With the development of advanced nuclear reactors such as fusion reactors and Generation IV reactors, structural materials for nuclear applications will face increasingly extreme service environments, including higher temperatures and stronger neutron irradiation. Existing structural materials may not be capable of meeting the performance requirements, emphasizing the necessity for developing new materials with enhanced high-temperature mechanical properties and irradiation resistance. High-entropy alloys (HEAs) have garnered significant attention in the field of nuclear materials due to their exceptional overall performance, particularly in the irradiation resistance. However, the composition of commonly HEAs based on face-centered cubic (FCC), body-centered cubic (BCC), and FCC + BCC phases, as defined by metallic systems containing four, five or more components in equimolar or near-equimolar ratios, does not meet the requirement of reduced activation. Thus, there is an urgent need to develop reduced-activation high entropy alloys (RAHEAs) for using in nuclear reactors. This article presents a review of the current research on RAHEAs, focusing on composition design, phase composition, mechanical properties, and irradiation properties. Among the various RAHEAs, two specific types deserve particular attention: nanoprecipitation-reinforced RAHEAs and oxide dispersion-strengthened (ODS) RAHEAs. These novel alloys may exhibit high mechanical properties and robust irradiation resistance, while also mitigating the brittleness issue often encountered with current RAHEAs. With a vast compositional space available for adjustments, RAHEAs hold significant potential for designing superior structural materials for use in advanced nuclear reactors. • It comprehensively reviews current research on reduced-activation high-entropy alloys (RAHEAs) for nuclear applications. • The review focuses on composition design, phase composition, mechanical properties, and irradiation properties of RAHEAs. • Two RAHEAs exhibit high mechanical properties and radiation resistance, while addressing brittleness concerns. • The extensive compositional space of RAHEAs offers significant potential for designing superior structural materials. [ABSTRACT FROM AUTHOR]

• First review on radiation safety issue of activated tungsten dust for fusion reactor. • Source terms of activated tungsten dust were detailed. • Tungsten environmental migration and transformation behaviors were illustrated. • Distribution and health effects of tungsten in human body were reviewed. • R&D suggestions for radiation safety assessment of activated tungsten dust were proposed. 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. [ABSTRACT FROM AUTHOR]

ITER machine availability is crucial for fusion reactors in order to be the main source of electricity, which is affected by many aspects. This paper focuses on maintenance aspects of the machine. As the plasma facing components inside the fusion reactors become highly activated caused by the neutrons generated by fusion reactions, maintenance of the fusion reactors needs to be done by remote means. In order to increase machine availability, it is required to have efficient maintenance scenario in terms of time and effort, which can be achieved through effective in-vessel remote handling (RH) operations, simplified maintenance of the RH system itself, well-coordinated logistics, and trained personnel, etc. All these aspects need to be investigated during the design phases, which requires a well-defined concept of operation of the RH systems. This paper describes the concept of operation of the Agile Robot Transporter (ART), an ITER in-vessel RH system, that is under development to carry out neutron and optical diagnostic calibrations, and removal/replacement of the blanket first wall samples. Based on the functional analysis of the required in-vessel tasks, system requirements are defined. Among them, key requirements to improve RH operation efficiency are identified and the ART is designed accordingly. Efficient deployment and removal operations of the ART are crucial so that it requires less vacuum vessel (VV) port opening and less cask transportation, which reduces needs for hands-on operations resulting in less human radiation exposure. Various and flexible tool exchange and storage concepts are proposed that allow parallel operations between the in-vessel and ex-vessel RH systems. Modular and re-configurable design is proposed for easy RH system maintenance, and effective rescue operations to rescue the RH system when it fails. • Concept of operation of a remote handling (RH) system for ITER vacuum vessel. • Remote handling system working in high radiation environment. • Multi degree of freedom heavy load robotic arm. • Efficient in-vessel component and tool exchange. • Rescue of the failed RH system. [ABSTRACT FROM AUTHOR]

• Plasma neutralisers can significantly improve the global efficiency of a negative ion based neutral beam injector. • Operation of a neutral beam injector on a fusion reactor imposes many constraints on the design of the injector and the plasma neutraliser, and materials compatible with the expected radiation environment, gas flow into the injector, and space constraints are very important for the design and operation of the plasma neutraliser. • The gas flowing into an injector is stored locally on the cryopumps. The total amount of D2 gas stored in the injector is strictly limited by the possible explosion between D 2 and O 2 if O 2 is present in the injector because of a leak from the surrounding atmosphere. The gas storage imposes periodic regeneration of the cryopumps. Considerations of the regeneration cycle imposes limits on the gas flow into the plasma neutraliser. • A beam driven plasma neutraliser has many advantages over a neutraliser using alternative methods to create the plasma in the neutraliser. If, due to the recombination of electrons and molecular ions, an acceptable plasma density not be achieved with D 2 as the target gas in the neutraliser, the use of argon as the target gas should be considered. The global efficiency of neutral beam injectors to be used on a fusion reactor needs to be of the order of, or greater than, 60 %, which is considerably higher than the efficiency of presently operating NBIs. The required efficiency can only be achieved by increasing the neutralisation efficiency from the ≈56 % obtained with the gas neutralisers currently used, which can be done using a plasma or photon neutraliser. This paper examines possible designs of a plasma neutraliser taking into account the many constraints imposed by operation in the environment around a fusion reactor, and a schematic design is proposed that meets those constraints. The design chosen uses a beam driven plasma neutraliser, which requires no additional power systems, diagnostics or interlocks. Using a plasma neutraliser is shown to improve the global efficiency to ≈55 %. Although that is below the desired value, reasons why that might be acceptable are discussed. This paper also discusses the possible use and benefits of such a neutraliser as an upgrade of the ITER heating neutral beam system. [ABSTRACT FROM AUTHOR]

• Compare national regulations on waste classification. • Ensure that there is no high level waste produced by EU DEMO. • Identify ways to improve the waste classification results (impact of material impurities, design, post processing of the waste...). • Identify the R&D perspectives associated to waste management. One of the top-level safety objectives for EU DEMO design and operation is to protect workers, the public and the environment from harm and thus to minimize radioactive waste hazards and volumes and ensure that the legacy to the future generation is limited. The objectives of the waste management strategy studies are to identify any showstoppers arising from the waste management or optimizations that could impact the design and operation of the facility and identify the R&D needed to ensure a safe management of these materials and waste. The following aspects are considered at this stage to anticipate and facilitate the waste management in particular with regards to waste that would require deep final disposal: limitation at source by selection of materials and impurities limiting the amount and toxicity of the waste, characterization of the fluxes, development of processes to reduce the waste radiotoxicity and amounts, definition of management routes and interim storage of waste. [ABSTRACT FROM AUTHOR]

Solid breeder materials for fusion power plants are subjected to extreme conditions of radiation damage, high temperature and designed component burn-up. The loss of lattice stability of Li 2 TiO 3 due to Li burn-up was investigated using first principles-based approaches to aid evaluation of upper limit of component lifetime. The lattice stability is analyzed using calculated phonon dispersion in the Li 2 TiO 3 supercell, while Li burn-up is modeled by the introduction of Li-vacancies. It has been determined that the studied ceramic can be structurally stable up to ∼40% of Li atoms burn-up. At 50% burn-up, the negative frequency branch in the phonon dispersion spectrum appears indicating the loss of lattice dynamical stability. Moreover, the structure obtained by introduction of unstable frozen phonons with minima energy amplitude also results in a phonon dispersion with negative branches in the structure obtained. Therefore, the system completely loses stability when approximately half of Li atoms are consumed. [Display omitted] [ABSTRACT FROM AUTHOR]

The vacuum extraction method with a mass spectrometry registration of tritium is presented in paper. It can provide a full-range analysis of gas phase composition in the chamber with samples under neutron irradiation. Lithium ceramics Li 2 TiO 3 in the form of pebbles with lithium enrichment on 6Li isotope of 96% was examined. The paper shows the results of reactor experiments on study the extraction of tritium-containing molecules from lithium ceramics Li 2 TiO 3 at various temperatures and reactor power levels at the WWR-K research reactor (Almaty, Kazakhstan). The time dependences of tritium yield from the ampoule with ceramics during reactor irradiation were obtained. Near-surface concentrations of tritium and hydrogen atoms on the ceramics surface under reactor irradiation were estimated. • The vacuum extraction method under irradiation at the WWR-K reactor was created. • The experiments were carried out with lithium ceramics samples. • In experiments, the tritium yield occurs mainly in the form of HT molecules. • The research will be continued. [ABSTRACT FROM AUTHOR]

Neutron decay time is an important parameter in physics, for which science still does not have a single definitive answer. Free neutrons decay into a pro-ton, electron, and antineutrino; however, different experiments give different results for the time a neutron takes to decay. The bottle method results in a neutron half-life of 14.64 minutes while the beam method results in a neutron decay time of 14.8 minutes. The third method, the standard model, results in a prediction of around 14.75 minutes. The experiment measured neutron decay using a cloud chamber and neutrons produced by a thermonuclear fusion using a fusor. In this case however given the lack of current application research, having a human classify events in the cloud chamber isn't effective, and where the application of Neural Networks and the application of cognition systems can be applied to make the task more accurate. In this paper, a method to directly measure neutron decay time based on the timing of cloud chamber events is presented, instead of calculating it from half-life results as other experiments have done this presented the details of training machine learning models for the application in neural networks and analysis using the cognitive system, Independent Core Observer Model (ICOM) to analyze results. [ABSTRACT FROM AUTHOR]

Attaining an amorphous crystalline structure imparts unique properties to certain metallic alloys that are impossible to achieve with conventional crystalline alloys. This current study investigates the amorphization potential of small additions of Zr, Ni, or V in Cu Ti alloys produced using the melt-spinning technique. The alloys are characterized using X-ray diffraction (XRD), differential thermal analysis (DTA), differential scanning calorimetry (DSC), and transmission electron microscopy (TEM) techniques. The need for filler materials with flexible and adaptable properties for joining the first wall and divertor components of future fusion reactors can be addressed by producing filler ribbons with amorphous structures. The results indicate that a combination of Zr (12 at. %) and Ni (8 at. %), when used as alloying elements with appropriate processing parameters such as a rectangular nozzle of 0.5 × 8.8 mm2 and a linear wheel speed of 30 m/s, can produce an amorphous structure. Only isolated nanocrystalline phases were detected by TEM analysis. As a result, ribbons fabricated under these conditions were several times longer and wider than those made using the Cu Ti master alloy. More importantly, they exhibit enough flexibility to conform to and cover a simulated pipe of the reactor. Conversely, the addition of V led to the formation of brittle phases. Ribbons fabricated under the optimal conditions were tested as filler materials in W-CuCrZr joints at 960 °C and were microstructurally and mechanically characterized, demonstrating their suitability for the intended application. • The combination of the melt-spinning technique and amorphizing materials avoid the formation of a crystalline structure. • The best results are achieved with a Cu20Ti12Zr8Ni at% alloy composition, a wheel speed of 30 m/s, with a rectangular nozzle measuring 0.5 x 8.8 mm². • Employing these processing parameters and alloy composition, ribbons measuring 70 mm in width and 160 mm in length are obtained. • The inclusion of V in the alloy results in embrittlement, preventing the samples from exhibiting sufficient flexibility. • Joints exhibit metallic continuity, and the microstructure comprises a copper-rich phase along with precipitates of Cu-Ti-Ni-Zr composition. [ABSTRACT FROM AUTHOR]

• The impact of the gamma irradiation environment in a fusion reactor on the electronic systems was introduced, and experimental verification was conducted through long-term Co-60 gamma irradiation experiments. • Gamma irradiation provides valuable reference for the design of electronic systems in fusion reactors. • Verified the operational condition of some chips under a long-term gamma irradiation environment (2000 Gy). • Provides reference for the verification of radiation resistance of electronic systems in a fusion reactor environment. In the environment of a fusion reactor, electronic systems are subjected to various influences including vibration, electromagnetic pulses, as well as irradiation from neutrons and gamma rays. Diagnostics systems on fusion reactors can accumulate radiation doses to a level where issues such as signal attenuation and system interruption can occur. Additionally, maintenance of electronic systems during fusion reactor experiments is often not feasible. Therefore, assessing the radiation tolerance of electronic systems used in fusion reactors is of paramount importance. To validate the radiation resistance of front-end electronics and cable in various diagnostic systems, a test circuit board was designed. This board integrates a variety of commonly used amplification chips and power supply chips to assess their performance in the environment of a fusion reactor. The experiment selected three different types of cables commonly used in fusion reactors for separate testing, to assess their parameter changes before and after irradiation experiments. Gamma Irradiation Experiments were conducted using a cobalt-60 (Co-60) radiation source at the Irradiation Center of Nanjing University of Aeronautics and Astronautics. The test circuit board was exposed to continuous gamma irradiation at two test points, with dose rates of 6 Gy/min and 1 Gy/min, respectively, for a total of 6.5 h, Data from the irradiation experiments were collected and analyzed. The results show that under continuous gamma irradiation, CJ7805 and AMS1117–5 chips stopped working at 586 Gy and 776 Gy, respectively. The output of the LT1175I5 chip dropped from -5 V to -1 V at 447 Gy. Apart from the radiation-resistant power supply chips, the output of other power supply chips slightly decreased. The signal reference of the charge amplifier dropped at 741 Gy but then slowly rose again, and the signal amplitude decreased at 900 Gy. The signal reference of the current amplification circuit remained unchanged, with a slight decrease in amplitude. The signal reference of the voltage amplifier exhibited jitter, with no change in amplitude. The experiment provides valuable data support for the radiation-resistant design of electronic systems in the environment of a nuclear fusion reactor. [ABSTRACT FROM AUTHOR]

The Bidirectional Reflectance Distribution Function (BRDF) of beryllium is important for a ray-tracing simulation to evaluate the effect of reflected light on optical diagnostics in the fusion device ITER. However, the BRDF for beryllium has not been investigated yet and is currently lacking. To fill this gap, this study measured BRDFs of beryllium specimens with different topographies using goniophotometers and optimized optical reflection models. Additionally, various surface topographic parameters were computed, and their correlation with parameters for the optical reflection models was explored. As a result, it was found that topographic parameters related to slope offer a certain correlation with the parameters of the reflection model. This can be useful for determining the optical reflection properties of materials for which direct BRDF measurements are not possible. • The Bidirectional Reflectance Distribution Functions (BRDFs) of beryllium are shown. • Topographic parameters have a certain correlation with a reflection model. • The maximum coefficient of determination for the correlation exceeds 0.91. • Improving the optical reflection model is the key to better correlations. [ABSTRACT FROM AUTHOR]

The brazeability of a cupronickel commercial alloy (Cu10Ni) was evaluated for its use as a filler alloy for high-temperature joining of tungsten to the reduced activation ferritic/martensitic steel EUROFER 97 (W-E) and between tungsten base materials (W-W) for its application at the first wall and divertor of future fusion reactors. In addition, given the importance of the residual stresses in these heterogenous joints, a study of the brazing conditions and the impact of the selected filler has been conducted using numerical software to understand its impact on the quality of the joint. Two thermal cycles were evaluated (1165 °C and 1190 °C) and selected based on the thermal characterization of the filler alloy. The microstructural examination revealed that, in W-E joints, nickel acts as an activator element, reacting and forming interfacial layers at the EUROFER 97 - Cu10Ni interface. In the case of the W-W joints, a lower level of diffusion phenomenon and metallurgical interaction between Cu10Ni and base materials were observed. The hardness profile indicated that the hardening process of EUROFER 97 was associated with the formation of untempered martensite. On the other hand, tungsten kept the received hardness. The mechanical characterization by shear test reported similar values between both types of joints carried out at 1190 °C but different when the temperature was increased (1165 °C), associated with the brittle character of tungsten and its lower metallurgical interaction. The numerical analysis of the brazing process carried out with ANSYS software shows that residual stresses are accumulated mainly at the interfaces. The information provided by the simulation shows, for a 50 µm filler thickness, the importance of mitigating the residual stress by selecting a filler with an intermediate Coefficient of Thermal Expansion (CTE) that alleviates mechanical stresses relative to the base materials. [Display omitted] [ABSTRACT FROM AUTHOR]

• Tritium production methods by loading Li rods into the HTGR has been developing. • All Li rods loaded in the HTGR have same design in previous elementary studies. • The performance of Li rods was evaluated by adjusting their arrangement and design. • Reducing the number of Li rods can be an effective optimization method. An initial tritium inventory is required to start fusion DEMOnstration Power Station (DEMO) reactors. However, a method to supply adequate tritium has not been determined yet. Tritium production via 6Li(n,α)T reaction by loading Li rods into the burnable poison (BP) holes of a high-temperature gas-cooled reactor (HTGR) has been proposed to address this problem (Matsuura et al., Nucl. Eng. Des. 243 (2012) 95 - 101). In previous preliminary studies, Li rods loaded in all BP holes were assumed to have the same design. This study evaluated whether the performance of Li rods can be improved for future optimization by adjusting the Li rod arrangement and the amount of Li compounds in them. The amount of tritium produced for gas turbine high-temperature reactor 300 (GTHTR300) was evaluated, while the total amount of Li compounds was maintained and the amount of loaded Li compounds changed depending on the layers and fuel regions. The maximum amount of tritium produced did not increase during the evaluations when reactor feasibility was satisfied. This implies that it is possible to reduce the number of Li rods while maintaining the amount of tritium produced for optimization, thereby reducing the costs of manufacturing Li rods and tritium recovery. GTHTR300 can produce 800 g of tritium in 360 days of operation using 2160Li rods. The results showed that the same amount of tritium could be produced by loading 720Li rods with the same number of fuel blocks. In addition, the effective multiplication factor, burn up, and power density of GTHTR300 were not significantly influenced during the operation. [ABSTRACT FROM AUTHOR]

In this work we present the latest progresses (September 2018) in the conceptual design of the main containment structures of DTT fusion reactor. The previous DTT baseline design is revised in terms of structural materials and overall reactor shape. The major change involves the vacuum vessel, which now foresees a welded double-wall stainless steel structure. The basic design includes eighteen sectors, with novel ports configuration for remote maintenance systems, diagnostics and heating equipment. New supports are designed for the first wall, which is conveniently segmented in view of assembly and remote replacement. The cryostat of the machine is conceived as a single-wall cylindrical vessel reinforced by ribs. The cryostat base is also in charge of supporting the vacuum vessel and the magnets system. A preliminary FEA analysis confirms that the main mechanical structure might withstand the design loads, in particular the ones resulting from possible plasma disruptions. [ABSTRACT FROM AUTHOR]

• An overview of few main design issues of DEMO reactor is outlined. • The possible solution for each of those issues has been briefly discussed. • The importance to develop in a homogeneous way the main systems of DEMO has been pointed in order to find an optimal solution to the issue. • The need of integration of design, safety in view of the nuclear license and the operation of the DEMO plant have been outlined. The EU DEMO preliminary safety and operating design requirements are being defined aiming at obtaining the nuclear license. The DEMO design approach is being organized, by taking into account the Nuclear Power Plant, ITER and Generation IV lessons learnt. Outstanding challenges remain in areas exhibiting large gaps beyond ITER. Those require a pragmatic approach, especially to evaluate and improve the readiness of technical solutions through dedicated physics and technology R&D. Therefore, a system engineering approach is adopted based on an integral plant design analysis which follows clear defined goals as safety, availability and power provision to the grid. This ensures not only the identification of critical interfaces but also the margin of possible solutions and, moreover, the definition of target parameters for technical systems to be met in order to arrive at an optimal DEMO design. The overall DEMO plant design has to be strongly safety and operation-balance of plant (BoP) oriented. The paper describes a set of important aspects of safety and BoP that require early attention and a continuous reanalysis at any significant design change. This includes: (i) plasma instability and mitigation actions; (ii) radioactive inventory minimization and control; (iii) energy inventory control and mitigation; (iv) power conversion system challenges and possible simplification; (v) huge and complex electrical loads; (vi) layout of the tokamak building versus plant and safety criteria. Any effort to reduce the complexity of a Fusion Power Reactor through simplification and rationalization of the design of the main systems translates into a more robust plant configuration enlarging safety margins and operational thresholds. [ABSTRACT FROM AUTHOR]

This work describes integrated numerical modelling applied to DEMO discharges with tungsten wall and divertor, using the COREDIV code, which self-consistently solves 1D radial transport equations of plasma and impurities in the core region and 2D multi-fluid transport in the SOL. The model is self-consistent with respect to both the effects of impurities on the α-power level and the interaction between seeded (Ar) and intrinsic impurities (tungsten, helium). This work is to analyse the influence of the helium and hydrogen recycling on the He confinement time in DEMO reactor. Simulations have been done for argon seeding puff 5.0 × 10211/s. It is found that recycling coefficient of helium have strong influence on the He confinement time, which increases from 7.25 s to 26.9 s going from lowest to highest recycling coefficient, but it has small influence on the effective charge state and radiation in the SOL. Alpha power decreases only by about 20%, which is the effect of main plasma dilution. [ABSTRACT FROM AUTHOR]

• The WCCB blanket is designed based on the array of the cooling tube assemblies and the rib. • The blanket module has three sets of independent coolant circulating systems. • The thermo-hydraulics scheme of the FW and the breeder zone is considered to be de-coupled. • The WCCB blanket may cover operation mode of 200 MW, 500 MW, 1 G W, and 1.5 G W fusion power. The water-cooled ceramic breeder (WCCB) blanket is one candidate for the Chinese Fusion Engineering Testing Reactor (CFETR), and it is being developed at the Institute of Plasma Physics of the Chinese Academy of Sciences. This paper presents an updated WCCB blanket concept design that is expected to cover the CFETR operation modes of 200 MW, 500 MW, 1 GW, and 1.5 GW fusion power and achieve tritium self-sufficiency with the latest core parameters (major radius R = 7.2 m; minor radius a = 2.2 m) and mission. The feasibility of the new blanket design is evaluated considering neutronics and thermal-hydraulics aspects. [ABSTRACT FROM AUTHOR]

• Systems codes assess the viability of fusion reactor designs using simplified models. • We have benchmarked SPECTRE and PROCESS using an SST-2 design for comparison. • Similar results are found, however with different plasma densities and temperatures. • Differences are caused by the treatment of radiation in relation to the L-H threshold. • We have performed Monte-Carlo based Uncertainty Quantification using PROCESS. Systems codes assess the viability of fusion reactor designs by using simplified models for the entire reactor system that allow for the exploration of large areas of parameter space. However, every design will have an associated uncertainty that arises from the accuracy of the models used, the assumptions made and the values of input parameters adopted. For individual codes, the uncertainty on their results can be quantified by investigating the dependence on the combination of input parameters. More generally, different codes can be compared against each other to test the underlying models. In this paper we compare the results of two systems codes, spectre and process , using a conceptual design for the SST-2 fusion reactor for benchmarking. We find that overall both codes produce similar results, however different plasma temperatures and densities are found due to the treatment of radiation in relation to the L-H threshold, and different fusion reaction rates. We then apply a Monte-Carlo based uncertainty quantification tool using process to find that while the design can produce in excess of 100 MW of fusion power, it is unlikely to produce pulses over 400 s. This is in agreement with previous work and suggests a larger aspect ratio is required. [ABSTRACT FROM AUTHOR]

Tritium retention in the vacuum vessel emerged as a potentially serious constraint in the operation of the International Thermonuclear Experimental Reactor (ITER) several years ago. Present-day outlook reveals that the main tritium amount will be retained in deposits in the divertor area and by repeatedly applied proper thermal treatment its amount would never exceed the permitted critical mass. During such heat treatments, the vacuum vessel will be heated, too, and it may represent a tritium sink, which is highly undesirable, as the kinetics of the tritium absorption rate is not known. The presented results on permeation through AISI 316 ITER grade stainless steel membranes in the pressure range between 50 and 1000 mbar at temperatures from 200 °C to 400 °C match well with reported values on AISI 316 steels. The accent is on the influence of the residual atmosphere on achieved permeability during heat treatments at moderate temperatures. [ABSTRACT FROM AUTHOR]

Abstract The European DEMO fusion power reactor (EU-DEMO) is still in the pre-conceptual design phase. The design strategy for the EU-DEMO hinges on investigating multiple reactor designs and technologies in parallel, progressively down-selecting these in the mid-2020s, in preparation for the conceptual design phase. The present implementation of the strategy centres around a baseline single-null design, which is configuration controlled and iterated approximately every two years. The majority of resources are dedicated to studying the baseline design; changing it is expensive, and takes months to do. Competing technologies for different sub-systems (e.g. blankets, magnets) are forced to co-exist within the same physical confines (CAD models), and conform to the same set of performance criteria. Meanwhile, the alternative reactor designs are only loosely defined, with no agreed set of parameters, no CAD models, and no formal framework for study. We argue that the EU-DEMO design strategy is best served by a more efficient implementation methodology which would enable more comprehensive exploration of the DEMO design space. We make the case for a change in approach to the design of DEMO-class reactors in Europe, and propose a solution which bears some resemblance to the present methodology, yet automates and accelerates the sequence of design and analysis activities when generating a design configuration. We present our preliminary attempts to demonstrate the feasibility of our idea, in the form of a new fusion reactor design code: BLUEPRINT. We demonstrate that the typical activities required to generate a DEMO design point can be sped up by four orders of magnitude —from months to minutes—paving the way for a rigorous and broad exploration of the design space. [ABSTRACT FROM AUTHOR]

Abstract From previous catastrophic fission nuclear accidents, such as the Chernobyl and Fukushima accidents, researchers learnt the lessons that external hazard beyond design basis or human errors could result in severe accidents and multi-failure of the confinements although they were considered as very-low-probability events and not requested to be paid much attention to according to the current nuclear safety regulations. Fusion energy is always regarded as a safe and clean energy. However, massive quantity of radioactivity still exists in the fusion reactor and is possible to be released into the environment. The environmental pollution and potential public consequences due to severe accidents of fusion reactor remain largely unexplored. In this contribution, we intended to investigate the hypothetical accident to envelop the worst but probable consequences of fusion reactor, and compare with historic Chernobyl and Fukushima accidents under assumed environmental conditions. It was demonstrated that, the radiation consequences of a hypothetical fusion accident would be much less severe than fission accidents, e.g. an INES 7 accident could not appear in a fusion reactor, as in the Chernobyl and Fukushima nuclear accidents. However, it would still be disastrous and the publics close to site might be exposed to "potentially lethal" radiation dose. Graphical abstract Image 1 Highlights • Consequences of fusion hypothetical accident were compared with fission accidents. • Contents includes source term, INES, dose, health, area pollution and restoration. • Release activity and dose factor of fusion are lower than Chernobyl and Fukushima. • Fusion may happen level 6 accidents but never level 7 like Chernobyl and Fukushima. • Environment restoration of fusion is much faster than Chernobyl and Fukushima. Using historic INES 7 nuclear accidents as baseline, this study firstly evaluates the comprehensive environmental pollution and public radiation consequences under a fusion reactor hypothetical accident. [ABSTRACT FROM AUTHOR]

Highlights • A regional angular adaptive algorithm to reduce angular discretization errors for shielding calculations. • The accuracy of this algorithm with FENDL/MG-3.0 libraries was investigated through analyses of streaming experiments with deuterium–tritium neutron sources. • The ARES results show good agreement with experimental values, for this narrow and deep gap structure, the calculation accuracy for the streaming analyses is mostly within ± 25% for fast neutron reaction rates. Abstract Shielding analyses for fusion reactors pose significant computational challenges, including strong anisotropic scattering and a combination of deep-penetration shielding and complex channeling. The discrete ordinates method is one of the most widely used techniques to solve the linear Boltzmann equation. We propose a regional angular adaptive algorithm to reduce angular discretization errors for shielding calculations. Angular adaptive refinement techniques are based on estimated local errors. The local errors are weighted by the importance distribution toward the computational goal, providing an appropriate goal-oriented adaptivity. The accuracy of this algorithm with FENDL/MG-3.0 libraries was investigated through analyses of streaming experiments with deuterium–tritium neutron sources at JAERI/FNS. From comparing reaction rates of measurements and calculations, the angular adaptive algorithm suffices in predicting the fast neutron field inside the gap. The solutions show that the algorithm can improve the accuracy of calculation results. [ABSTRACT FROM AUTHOR]

Abstract In this contribution, small-angle neutron scattering (SANS) studies of the micro-structural radiation damage in technical steels, developed for application in future fusion reactors, are presented. The effect of neutron irradiation at 250 °C–450 °C, up to dose levels of 16 dpa (displacement per atom), has been investigated in the European reference ferritic/martensitic steel Eurofer97 (0.12 C, 9 Cr, 0.2 V, 1.08 W wt%), both in its standard composition and mechanically alloyed with B (up to 1000 appm) to enhance the helium production, via transmutations, and reproduce fusion relevant helium/dpa ratios. The obtained SANS results suggest that in the irradiated standard Eurofer97 micro-voids distributions are present, with very small average radii (a few Å) and volume fractions (around 0.01); such low values, close to the SANS resolution limit, are consistent with the good resistance of this steel to radiation damage, for low helium concentrations. In the irradiated B-alloyed Eurofer97, more complex changes are observed: the helium bubble distributions appear strongly dependent on the irradiation temperature and on the helium concentration; furthermore, micro-structural effects possibly related to modifications in the steel matrix are detected. [ABSTRACT FROM AUTHOR]

Highlights • Large-sized W-Y 2 O 3 bulk material was successfully produced. • Rolling deformation induced anisotropic microstructure and fiber geometric structure. • Thermal conductivity has a linear relationship with reciprocal temperature. Abstract Tungsten (W) or W alloys are the main candidates for plasma-facing materials of future fusion reactors. Large-sized W-Y 2 O 3 bulk material has been successfully produced in this study for future engineering applications. Wet chemical method and continuous hydrogen (H 2) reduction were applied to achieve mass preparation of W-Y 2 O 3 composite powder. The final bulk material was obtained using the H 2 atmosphere sintering technique and rolling deformation process. Conventional X-ray diffraction patterns were used to characterize the different surfaces of the rolled specimen, which coincide with the results of electron backscatter diffraction detection, to evaluate the texture information. The 50% rolled W-Y 2 O 3 bulk material have three types of textures, namely, θ-fiber, α-fiber, and γ-fiber. Notably, the thermal conductivity exhibits no obvious difference between sintered bulk at the original state and rolled bulk along various directions. After rolling deformation of the W-Y 2 O 3 bulk material, the effect of the anisotropic microstructure on the thermal conductivity can be offset by the effect of the fiber geometric structure. For the body-centered cubic structure of W, the thermal conductivity versus the crystal orientation should be λ [100] > λ [110]. [ABSTRACT FROM AUTHOR]

• The preliminary PSA, items safety classification based on PSA, and reliability allocation based on PSA, were explored for fusion reactors. • The multi-objective optimization among the PSA and its applications were discussed for fusion reactors. The fusion reactors, which adopt Deuterium-Tritium (D-T) nuclear reaction to produce energy, are promising in the future. The nuclear safety of fusion reactors is an important research topic on the road of fusion reactor engineering. The Probabilistic Safety Assessment (PSA) is one of the important tools to evaluate the safety concerns of fusion reactors, and the importance obtained from the PSA of fusion reactors can provide references for the items safety classification and reliability allocation. Reasonable safety classification and reliability allocation are of great significance for improving the safety and economy of fusion reactors. The PSA, items safety classification and reliability allocation for fusion reactors influence each other, form a whole, and can be considered as a multi-objective optimization problem. In this contribution, preliminary PSA and its applications for fusion reactors, which include the preliminary PSA, items safety classification and reliability allocation based on PSA for fusion reactors, were performed. Then, the multi-objective optimization among the preliminary PSA and its applications for fusion reactors were preliminarily discussed. The results show that the proposed systematic approach can provide important references for the optimization and improvement of the safety and economy for fusion reactors. [ABSTRACT FROM AUTHOR]

The thermal fatigue effect on the microstructure and mechanical properties of the joints that form some components of the future fusion reactor is a concern within the scientific community. In this study, we analyze the metallurgical modifications caused by thermal fatigue and their impact on the mechanical properties of tungsten-EUROFER brazed joints (blocks measuring 6 × 6 × 4 mm). We conduct the analysis using an actively cooled mock-up subjected to steady-state thermal loads, which provides valuable information about the operating conditions of the reactor. Three different surface conditions of tungsten were evaluated: 600 ºC (2 MW/m2), 700 ºC (2.5 MW/m2), and 800 ºC (3 MW/m2), with varying numbers of applied cycles ranging from 100 to 1000. Throughout the tests, infrared cameras and pyrometers were used to analyze the thermal behavior of the W-EUROFER joint. At 600 ºC and 700 ºC target temperatures, no anomalies in the heating and cooling capacity of the W-EUROFER joint were observed. This represents an advancement compared to previous studies that employed Cu20Ti filler, as it demonstrates consistent and efficient cooling capabilities even at surface temperatures of up to 700 ºC, without any notable anomalies starting from the previous filler's 500 ºC. However, in the case of 800 ºC, the test had to be prematurely stopped. Microstructural analysis revealed the formation of cracks in some cases due to the stresses generated by the mismatch in the coefficient of thermal expansion between the materials used. These cracks affected the mechanical integrity of the joint. [ABSTRACT FROM AUTHOR]

• Demonstrated the utility of low-attenuation T(0,1) mode ultrasonic guided waves for long-distance omnidirectional defect localization in fusion reactor pipelines. • Established optimal frequencies for detecting minor and irregularly-shaped defects by experimental and numerical simulation. • Emphasized the dominance of diameter in circumferential defect localization. • Combined the cross-correlation and Hilbert algorithm to mitigate signal delay issues during axial defect localization. The stable operation of a fusion reactor depends on the reliable functioning of various accessory systems, many of which comprise extensive pipelines with different diameters and diverse cladding. Defects in pipelines can occur due to manufacture, thermal stress, and fatigue, threatening the systems' safety and reliability. Ultrasonic guided waves (UGWs) testing offers a promising approach for remotely locating these minute defects. The study explores non-dispersive fundamental torsional guided (T(0,1)) waves, particularly their influence on minor defect detection in stainless steel pipes. Both finite element simulations and experimental trials reveal that reflection coefficients, indicative of defect presence, rapidly escalate from 64 to 128 kHz. The optimal frequency for minor defect detection lies between 128 and 180 kHz. A significant finding from the circumferential distribution spectrum and angle profile reveals pipe diameter's dominant influence on the reflection energy distribution concentration. Furthermore, this highlights the unique ability of T(0,1) guided waves to precisely locate defects in the circumferential direction within large-diameter pipes over long distances. The results underscore the promising potential of UGWs testing in complex pipelines of fusion reactors. [ABSTRACT FROM AUTHOR]

In this paper, a theoretical model is established to describe pebble-crushing behavior, and the fractal theory was introduced to describe the relationship between original particles and fractures. Considering the size effect and random variation of crushing strength, the Weibull theory was used to characterize the strength variation, and the correlation between the Weibull modulus and fractal dimension was derived. The impact test was performed to verify the established theoretical model's validity. Results showed that the theoretical model was verified to be valid for crushing behavior description, and the slope of the fitted straight line could be used to calculate the fractal dimension, which is the most critical parameter in fractal theory and could describe the extent of crushing in the pebble bed. The crushing behavior of pebble beds under impact loading was also studied, and test results identified limited evidence of fractal dimension limitation as the impact times increased. The crushing degree could not increase by only increasing the impact times. The extent of crushing had a limitation at a certain drop height. The evolution rate of crushing behavior negatively correlated with total impact energy, and the extent of crushing changed more drastically in a single-layer bed. The established theoretical model and findings presented in this paper add to our understanding of granular material physics, especially the mechanism of particle crushing behavior in a packed bed for fusion blankets. • A theoretical model of pebble-crushing behavior in fusion blankets was established. • Fractal theory and Weibull distribution were introduced in the theoretical model. • A series of impact tests on the pebble bed were performed for model verification. • The fractal dimension had a limitation as impact energy kept increasing. • The crushing evolution rate changed more drastically in a single-layer bed. [ABSTRACT FROM AUTHOR]

• The long-term creep behavior of RAFM steel relevant to ITER-TBM operation conditions was investigated. • The average size of newly formed laves phase is not larger than that of M 23 C 6 carbides during the long-term creep process. • The creep void led by the large-size precipitates is the failure mechanism for the long-term creep. The reduced-activation ferritic-martensitic (RAFM) steel serves as the structural material for the blanket of the International Thermonuclear Experimental Reactor (ITER) and future fusion reactors. In order to achieve a design lifetime of 2–3 years for the blanket in the fusion reactor, it is imperative to thoroughly understand the long-term (>20,000 h) creep behavior of RAFM steel. In this regard, the present study investigated the long-term creep damage mechanism of RAFM steel under creep periods greater than 20 000 h. The results indicated that the creep exponent decreases as the creep rupture time increases from several thousand hours to more than 20,000 h. The microstructure analysis revealed that dislocation slip was effectively pinned by the precipitates in the long-term creep process. Additionally, the M 23 C 6 and Laves phase played a critical role in inhibiting interfacial slip and preventing lath coarsening. However, a few micron-scale M 23 C 6 and Laves phases led to the formation of cavities in the boundary. These cavities subsequently polymerized and aggravated into cracks, ultimately becoming the primary mechanism for the long-term creep fracture. [ABSTRACT FROM AUTHOR]

Highlights • Advanced technology modules were added to the Sycomore fusion reactor system code. • Increased maximum toroidal field coil steel stresses yield large reactor size reduction. • Increase of the tritium burn-up requires large reactors. • Water-cooled and helium-cooled divertors are equivalent for DEMO1-class pulsed designs. Abstract The next step for fusion energy after the ITER tokamak is the demonstration power plant DEMO. In this framework, system codes are used to address high-level key design issues for the DEMO pre-conceptual phase. They aim at capturing the interactions between the subsystems of a fusion reactor. SYCOMORE is a modular system code which includes physics and technology models coupled to an optimizer in order to explore a large design parameter space. In the present paper, trade-off studies focused on technology modules are reported including the influence of some design-driving assumptions on the reactor performances and size, starting from a European DEMO1-like design (more than 500 MW net electric power and 2 h burn duration). The increase of the mechanical stress limits in TF and CS magnets can help reducing the reactor size, slightly more when high temperature superconductors are used in the TF coil. The tritium breeding ratio can be improved to more than 1.10 by a moderate increase of the size, but the tritium burn-up ratio needs one additional meter of major radius for every percent increase. Divertor coolant options are also compared, showing some differences between helium, hot and cold water scenarios at various incident divertor heat fluxes. [ABSTRACT FROM AUTHOR]

Highlights • The COREDIV code has been used to simulate DEMO inductive discharges with different impurity seeding (Ar, Xe) and mix gasses. • For xenon seeding for case with standard radial diffusion in SOL, such regime of operation seems not to be possible. • For Ar and combined Ar+Xe seeding it is possible to achieve H-mode plasma operation with acceptable level of the power to the target, but is upper limit on Xe concentration. • The influence of the prompt re-deposition model on the DEMO working point is relatively weak. Abstract In this paper numerical simulations with COREDIV code, which self-consistently solves radial transport equations in the core region and 2D multi-fluid transport in the SOL of EU DEMO discharges in full tungsten environment (W divertor and wall) for H-mode scenario has been performed with different impurity seeding. The optimal impurity mix for reactor would require the use of high Z impurity (low dilution, higher P α) in core and low/medium Z impurity radiating in the SOL region. In this paper we focus on investigations how the operational domain of EU DEMO can be influenced by different seeding (Ar, Xe) and mix gasses. In case of Xenon seeding, H-mode operation can NOT be achieved for standard radial diffusion (D SOL = 0.5m2/s) in the SOL. In the simulation, we have considered combined seeding of Ar (good radiator in SOL) and Xe (good radiator in core). In order to find the optimal impurity mix, simulations have been performed in such a way that for a few fixed levels of Xe puff, Ar puff has been increased from zero to the maximum value allowed by the code stability. For combined seeding: xenon + argon working point for EU DEMO can be found. [ABSTRACT FROM AUTHOR]

Highlights • Design of a compact and cost-efficient helical fusion reactor FFHR-c1 is given. • FFHR-c1 is equipped with the liquid metal divertor and the cartridge blankets. • The liquid metal divertor is maintained basically with simple up/down motions. • The cartridge blanket is maintained with a simple 4DOF motion at a fixed angle. • The availability of FFHR-c1 can be larger than 70%. Abstract The maintenance scheme in the compact helical fusion reactor, FFHR-c1, equipped with the liquid metal divertor, REVOLVER-D (Reactor-oriented Effectively VOLumetric VERtical Divertor) and the cartridge-type molten salt blankets, CARDISTRY-B (CARtridges Divided and InSerTed RadiallY − Blanket), is investigated. The magnetic configuration of the compact variant FFHR-c1 is similar to that of the Large Helical Device (LHD), while the device size is 2.8 times enlarged from LHD and a strong magnetic field strength of ∼8 T at the plasma center is adopted. The maintenance of the REVOLVER-D is simpler than that of the helical divertor with a complicated structure as seen in the LHD. In the REVOLVER-D, showers of molten tin are injected into the ergodic layer at 10 inner ports. To circulate the molten tin, 10 sets of the shower system including a liquid metal pump, ducts, a showerhead, a pool, and a heat exchanger, are installed. These can be replaced with simple up/down motions. The CARDISTRY-B consists of 320 tritium breeding blanket cartridges, which are toroidally segmented every two degrees. These cartridges are maintained by using heavy manipulators with a simple 4DOF motion at a constant toroidal angle. [ABSTRACT FROM AUTHOR]

Abstract The requirements for Japanese DEMO are to demonstrate (1) steady and stable electrical power generation, (2) plant availability on a level relevant for commercialization, and (3) self-sufficiency of fuels. The DEMO design activity has been mainly focused on challenging DEMO design issues such as divertor heat removal, superconducting magnets, breeding blankets, remote maintenance and the water cooling system, in order to reach the systemic consistency toward a promising DEMO concept. The current status of the DEMO conceptual design activity is presented. [ABSTRACT FROM AUTHOR]

Highlights • Tritium production using high temperature gas-cooled reactor (HTGR) is considered. • An irradiation test on High Temperature engineering Test Reactor (HTTR) is assumed. • The Li rod suited to the HTTR and produces 30 g of tritium in a year was designed. • The scenario of the irradiation test to confirm the Li rod performance was presented. • Li rod irradiation test to demonstrate the tritium production in the HTGR was proposed. Abstract Large quantities of tritium are required to start up fusion reactors and conducts engineering tests using tritium for a fusion blanket system. However, tritium is very rare and kg orders of tritium must be produced artificially. Tritium production, by 6Li(n,α)T reaction using a high temperature gas-cooled reactor (HTGR) has been proposed. This method considers the loading of Li rods into burnable poison holes in the HTGR. In this paper, the Li rod suited for use in the High Temperature engineering Test Reactor (HTTR) was designed, and tritium production and leakage from Li-rod capsules were evaluated by adjusting the thicknesses of LiAlO 2 , alumina, and Zr layers. An irradiation test scenario to be conducted in the HTTR for demonstration of the Li rod's tritium production and containment performance was presented. [ABSTRACT FROM AUTHOR]

Abstract Thermal and flow analyses on the cartridge-type molten salt blanket named the CARDISTRY-B (CARtridges Divided and InSerTed RadiallY – Blanket) for the compact helical fusion reactor FFHR-c1 have been carried out. As a result, it is shown that the maximum local temperature in the cartridges can be kept below 550 °C, which is an allowable temperature of the structure material, under the neutron wall loading condition of FFHR-c1. The CARDISTRY-B is composed of the tritium breeder (Breeding Blanket: BB) and the neutron shield (Shielding Blanket: SB). Both of them are toroidally segmented every two degrees. From a safety perspective, the molten salt FLiNaBe (melting point ∼305 °C) is chosen as the first option of the breeder material. In the case of FLiNaBe BB cartridge made of RAFM (Reduced Activation Ferritic/Martensitic steel), it is necessary to suppress the maximum temperature of BB cartridges to 550 °C or below. In this study, a finite volume method code considering beryllium (Be) pebbles inside a flow channel has been used to evaluate the cooling characteristics of a BB cartridge. By estimating the pressure drop through pebbles with the porous medium model, we have investigated how the temperature of BB cartridge change depends on the mass flow rate of FLiNaBe. The volumetric heating profile was modeled according to the neutronics calculation results obtained by the MCNP (Monte Carlo Neutron and Photon Transport) code. [ABSTRACT FROM AUTHOR]

Space allocation is important in the design of a fusion reactor, where numerous components operate under complex conditions, including a strong magnetic field, high-energy neutron irradiation, and a wide temperature range (4.5 K at the magnet and ∼300–500℃ at the blanket). To ensure the consistency of the China Fusion Engineering Test Reactor (CFETR) design, multi-component interference checks and distance measurements must be routinely performed. In this paper, a space analysis module is developed on the CFETR Integration Design Platform to provide the space analysis function for both rigid and deformed components (using the Representative-Triangles method). The example of the space allocation of a vacuum vessel (VV) and a thermal shield (TS) is demonstrated. The deformation due to heating under the bake-out condition is considered (during operation, heating and swelling due to radiation also cause deformation). With the input from the component design module, the space analysis can be well-performed for both rigid models and thermal deformation results (under the simplified boundary condition). Through iteration between the component design modules and the space analysis module, a preliminary designed clearance of 40 mm between the VV and TS is insufficient owing to the deformation during bake out, and the optimized clearance increases by a factor of two. [ABSTRACT FROM AUTHOR]

The interfacial characterization of W-Eurofer brazed joints, which could be implemented in the first wall component of the future DEMO nuclear fusion reactor, has been carried out by TEM and nanoindentation. The Eurofer-braze interface is characterized by the formation of two reaction layers formed during the melting stage of the filler in the brazing process, which reported high hardness and modulus values. The obtained microstructure supposes a homogenous transition between the base material and the braze. However, in the W-braze interface a more defined interface is formed, where the presence of an intermetallic transition compound has been identified. The combination of both techniques allows to study the resultant microstructure and to analyze the mechanical response of the phases, which constitute the interface and could be helpful to control the brazing conditions avoiding undesirable brazeability problems. [ABSTRACT FROM AUTHOR]

This paper is written in honor of my long term friend and mentor Professor Bernard Roth who spent over six decades in Academic contributions devoted to Kinematics, Dynamics, Control, and design of computer controlled mechanical devices. He is continuing actively in his work from the famous school Hasso Plattner Institute of Design cofounded at Stanford (popularly known as d.school). As an Academic Director he is active in Creativity in Design. This paper explores the idea of Creativity directly from Science of 17th–18th centuries in place of Engineering evolved by Professor Timoshenko at Stanford in early 20th century at the time of rapid expansion of rotating machinery from de Laval in 1882. An example of design that involves extreme temperature ranges, magneto hydrodynamics and optimization in a fusion reactor to produce tritium fuel will be described. [ABSTRACT FROM AUTHOR]

Blanket is implemented in fusion reactor for removing heat, shielding neutrons, and tritium breeding. Under accident conditions, natural circulation flow will emerge due to the buoyancy forces caused by density gradient in the different parts of the loop. Without the function of moving machinery, the natural circulation will occur to remove the heat of the nuclear reactor core, which denotes natural circulation is one of the significant passive cooling mechanisms of a new reactor. The appearance of subcooled boiling will greatly affect density gradients in the closed loop and lead to the flow instability of coolant. Experimental study has been performed to investigate natural circulation subcooled boiling characteristics in square channels of fusion blanket. Experimental results show that the natural circulation mass flow ascends with the increase of the heat flux and the decline of pressure, and subcooled boiling is easier to appear in natural circulation because natural circulation flow is usually small and heat transfer intensity is lower compared to forced circulation. The location of onset of nucleate boiling (ONB) moves to inlet of test section and the wall superheat increases with the ascend of heat flux. A new empirical experimental correlation is developed on the foundation of experimental data, which correlates heat flux at ONB in natural circulation with the maximum relative error of 35%. Furthermore, the correlations of wall superheat, location of ONB, and subcooled boiling heat transfer coefficient are also acquired to predict the occurrences of natural circulation subcooled boiling of square channel. [ABSTRACT FROM AUTHOR]

Design of a fusion reactor power plant requires putting together physics elements (plasma physics) and engineering elements (plant characteristics). System codes are tools that perform integrated plant design taking into account those two aspects. Design optimization can also be carried out, to produce the “best” reactor design, according to some pre-defined figures of merit. However, presently used system codes often lack a plasma model with sufficient level of realism in terms of description of plasma processes and non-linearities. In this work, for the first time, a framework in which the plasma model should interact with the engineering elements is detailed, giving the attention on the logic of how the plasma physics should be represented inside the system codes. Ultimately, no ambiguity is left on how the problem should be addressed and solved. Concrete details on how the plasma model should be written are also presented. A novel code PLASMOD has been written which incorporates these elements and can be used in a generic engineering system code. [ABSTRACT FROM AUTHOR]

Hydrogen isotopes retention and bubbles formation are critical issues for tungsten as plasma-facing material in future fusion reactors. In this work, the formation and growing up behavior of hydrogen bubbles in tungsten were investigated experimentally. The planar TEM samples were implanted by 6.0 keV hydrogens to a fluence of 3.38 × 10 18 H ⋅ c m − 2 at room temperature, and well-defined hydrogen bubbles were observed by TEM. It was demonstrated that hydrogen bubbles formed when exposed to a fluence of 1.5 × 10 18 H ⋅ c m − 2 , and the hydrogen bubbles grew up with the implantation fluence. In addition, the bubbles' size appeared larger with higher beam flux until saturated at a certain flux, even though the total fluence was kept the same. Finally, in order to understand the thermal annealing effect on the bubbles behavior, hydrogen-implanted samples were annealed at 400, 600, 800, and 1000 °C for 3 h. It was obvious that hydrogen bubbles' morphology changed at temperatures higher than 800 °C. [ABSTRACT FROM AUTHOR]

The compound Nb 3 Sn possess superconductivity at suitable temperatures, therefore, it is best suited to be used in the toroidal coils of superconducting magnets which holds the fusion plasma and confine it inside the reactor core. The neutron induced reaction cross-sections are required from threshold to 20 MeV for different isotopes of Tin (Sn). Since limited data is available for the reactions with the Sn isotopes. Therefore, we have optimized the (n, p) and (n, 2 n) reaction cross-sections for all possible Sn isotopes from threshold to 20 MeV with modified input parameters in the nuclear reaction modular codes EMPIRE-3.2.2 and TALYS-1.8. These codes account for the major nuclear reaction mechanisms, including direct, pre-equilibrium, and compound nucleus contributions. The present results from 116 Sn ( n , p ) 116 m In , 117 Sn ( n , p ) 117 In , 117 Sn ( n , p ) 117 m In , 118 Sn ( n , 2 n ) 117 m Sn , 120 Sn ( n , 2 n ) 119 m Sn and 124 Sn ( n , 2 n ) 123 m Sn reactions calculated with nuclear modular codes: TALYS – 1.8, EMPIRE – 3.2.2 were compared with EXFOR data, systematics proposed by several authors and with the existing evaluated nuclear data library ENDF/B-VII.1, as well. The results from the present study can be used for the future development of ITER devices as well as to upgrade the nuclear model codes. [ABSTRACT FROM AUTHOR]