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A discharge-type fusion neutron source generates neutrons by fusion reactions of hydrogen isotope atoms. In order to operate the fusion device based on the deuterium–tritium (D–T) fusion reaction, the tritium inventory is required to be decreased. In the present work, a self-sufficient system was installed into the fusion device for reducing the amount of hydrogen isotope fuel gas. The fuel gas was supplied and recovered with an intermetallic compound ZrCo in a sealed chamber in this system. A deuterium–deuterium operation was maintained for more than 60 min with stable discharge voltages, and the temperature of the ZrCo bed was changed to improve the neutron production rate. Factors that influenced the pressure inside the chamber were determined and optimized. Gas analysis using a quadrupole mass spectrometer indicates a dilution of the deuterium fuel was caused by hydrogen isotope exchange between the supplied deuterium gas and the absorbed protium on the chamber and electrodes surface. This system's hydrogen isotope control method and the amount of fuel were proposed for a D–T operation.

Understanding the permeation behavior of tritium from a pebble bed breeding blanket is essential for establishing a self-sufficient fuel cycle in a nuclear fusion reactor. It is known that double corrosion layers forms on reduced activation ferritic-martensitic (RAFM) steel surface by gas release from a ceramic breeder material, however, its effect on hydrogen permeation behavior has not been elucidated. Herein, in-situ measurement of hydrogen permeation through an F82H RAFM wall of a ceramic breeder pebble bed was performed under H₂-added sweep gas conditions. The corrosion layer formed on the F82H sample had a dense microstructure, which reduced hydrogen permeation flux at least by one order of magnitude. The permeation reduction factors were 20–50 at the water-coolant temperature of a blanket. A self-repairing ability is expected for the surface oxide layer as the corrosion occurs spontaneously inside a breeding blanket.

Since the low-carbon technology fusion energy is scheduled to be commercialized beyond 2050, the late-mover needs to flexibly and strategically operate for sustainable cash flow responding to the future market mechanism. Carbon sequestration of the fusion–biomass hybrid system by using low-temperature heat is proposed under the two preconditions: the establishment of the deregulated electricity market, and the emission trading market. When the price is above the breakeven price, the hybrid system generates electricity; otherwise, charcoal production is performed. Charcoal production is a stable solid form of carbon capture, and can possibly be traded on the emission credit market to secure profit. The experimental result shows that it is possible to produce charcoal at low temperature of 280 °C, indicating that any type of blanket design is applicable in the hybrid system concept. The carbon sequestration cost is 39 $/tCO2 at the default parameter of net electric output of 4000 $/kWe for the fusion power plant, and 60% of charcoal yield from experimental result. Comparison between the hybrid system and the fusion power plant proves that the fusion–biomass hybrid system prevents further loss, by setting two-track income structures in the various market mechanisms.

In 2013–2014, the FliNaK/LiPb twin-loop system Operational Recovery Of Separated Hydrogen and Heat Inquiry-2 (Oroshhi-2) was constructed at National Institute for Fusion Science as a liquid blanket...

This report introduces a current status of the continuous tritium recovery test campaign of PbLi droplets in vacuum, using vacuum sieve tray (VST). The campaign aims to verify the viability of VST method at a prototype design level. The following verification are to be performed. 1) Verify the steady state extraction efficiency of tritium using a 0.6 mm nozzle. The target is greater than 80%. 2) Verify the mutual interference effects on efficiency degradation by multiple droplets by comparing 1, 4, 7, and 19 nozzles. 3) Verify the reliability of VST in continuous operation. At least 24 h of non-stop operation without efficiency degradation is the primary goal, which can be extended up to 48 h. A dedicated extraction setup is to be installed March 2019 into the Oroshhi-2 PbLi experimental loop at the National Institute for Fusion Science (NIFS) Toki-Japan, followed by a stand-alone test with a short flowing period at Kyoto University which is already finished. Deuterium (D2), instead of tritium, is used for the experiment. Extraction efficiency is calculated by comparing the D2 concentration before-drop and after-drop. The amount of released D2 is also measured by a quadrupole mass spectrometer (QMS) and verified by the concentration difference. The PbLi temperature is between 375 ℃ and 450 ℃, and the flow velocity at the nozzle is between 1.5 m s−1 and 4.5 m s−1. The final report is scheduled to be completed by the end of 2019.

In this study, hydrogen permeation through Flinabe (equimolar LiF + NaF + BeF2 molten salt) including 0.1 wt% Ti particles is experimentally and analytically investigated. Ti particles mixed with Flinabe can suppress H2 permeation for a certain time when the upstream side is supplied with H2. The diffusion time lag after activation process is over 20 times larger than that before activation. H2 diffusion behavior in the Flinabe + Ti system is analyzed using an unsteady-state diffusion equation. The effective diffusivity in the Flinabe + Ti system is around 1/27 of that in the Flinabe system. It is confirmed that the activation of Ti particles is important for the overall permeation process in the Ti-Flinabe system.

We discussed with experiments the selective heating behavior of a mixture of molten salt and metal powder containing hydrogen in a single-mode cavity by 2.45 GHz microwave. The experimental device consists of two single-mode devices that intersected perpendicularly. In the common cavity at the two crossing single-mode devices, one microwave generates the maximum electric field of standing wave, and another microwave generates the maximum magnetic field of microwave. Thus, the sample located at the crossing can be selectively heated by the inputting electric field and/or the magnetic field. The frozen salt can be heated and melted by the electric field. Hydrogen was recovered from the metal powder in the molten salt by magnetic heating.

Sacrificial corrosion protection, a new use of Ti powder in molten salt blankets, is proposed. Redox reactions of mixed Ti powder into molten fluoride salts are predicted to suppress the TF corrosion of structural materials. Stainless Steel 430 (SS430) is immersed into FLiNaK containing HF at 823 K for 6 h. Before the immersion, the Ti powder either exposed to air or not exposed to air at all is mixed into FLiNaK. It has been found that either types of mixed Ti powder can suppress the HF corrosion of SS430.

The design of liquid metal cooling system for the liquid blanket concepts in magnetic fusion reactor requires the magnetohydrodynamics (MHD) friction loss coefficients for various pipes including pipes and bents to estimate the pressure drop of the whole blanket. The aim of this study is to reveal the MHD pressure drop of PbLi flow in a double-bended pipe. The experiments were conducted in the Oroshhi-2 loop having a super conducting magnet of 3T at NIFS in Japan. In this paper, the MHD pressure drops of the pipe under the magnetic field of 0.5T–3.0T, the ranges of Reynolds number (Re) of 2000–200,000 and Hartmann number (Ha) of 106–640 at the upstream, double-bended and downstream regions were measured. As the results, all the MHD friction loss coefficients in three regions were well correlated with the interaction parameter (Na = Ha2/Re) in the range of 0.1

We report a neutronics study of a blanket mock-up using a discharge-type compact fusion neutron source. Deuterium–deuterium fusion neutrons were irradiated to the mock-ups composed of tritium breeder and neutron reflector/moderator. The tritium production rate (TPR) per source neutron was measured by a single-crystal diamond detector with a 6Li-enriched lithium fluoride film convertor after the calibration process. Despite the low neutron yield, energetic alpha and triton particles via 6Li(n, t)α neutron capture as well as 12C via elastic scattering were successfully detected by the SDD with high signal to noise ratios. The TPRs were experimentally evaluated with errors of 8.4%–8.5% at the 1σ level at the positions with high thermal neutron fluxes where the errors were dominantly introduced by uncertainties in the monitoring of the neutron production rate. The calculated to experimental (C/E) values of TPR were evaluated to be 0.91–1.27 (FENDL-2.1) and 0.94–1.28 (FENDL-3.1). As the neutron source can generate 14 MeV neutrons using a mixed gas of deuterium and tritium, this approach provides more opportunities for blanket neutronics experiments.

FLiNaBe is a promising liquid blanket material for a fusion reactor. Since hydrogen solubility for FLiNaBe is low, one concern is the permeation of bred tritium. In order to increase effective solubility of hydrogen, the addition of Ti powder was proposed. In this study, the influence of Ti addition on the behavior of tritium generated inside FLiNaBe by neutron irradiation was investigated. The samples of FLiNaBe and FLiNaBe mixed with Ti of 5 wt% and 0.5 wt% were prepared and the solid-state samples were irradiated by thermal neutrons at Kyoto University Research Reactor. The neutron-irradiated samples were heated to 700 °C in an Ar gas flow. The released tritium was collected by a water bubbler and the temporal variation of TF and HT release was observed. Initially, the most of tritium was released as HT from both FLiNaBe with and without Ti but eventually, the most of released tritium was HT after long time heating from FLiNaBe with Ti. This result suggests that the addition of Ti can decrease the release of TF very much. The amount of tritium released from FLiNaBe with Ti was lower than that from FLiNaBe without Ti. This result suggests that the addition of Ti can suppress the release rate of tritium from FLiNaBe.

Liquid lithium, a candidate material for the target of International Fusion Material Irradiation Facility (IFMIF), easily contains nitrogen. Nitrogen in liquid lithium should be removed as it accelerates corrosion of tubing materials. Hot trapping by iron-titanium alloys is expected to be a practical method. In order to understand the characteristics of nitrogen trapping with iron-titanium alloys in liquid lithium, nitrogen distribution was observed by Soft X-ray Emission Spectroscopy (SXES). Samples of Fe-5Ti alloys were immersed for 9–256 h in liquid lithium which contains 1000 w-ppm of nitrogen at 823 K and studied by X-ray Powder diffraction and SXES. The results show that nitrogen is trapped mainly in the grain boundaries and on the surfaces of the Fe-5Ti alloys and diffuses in Fe-5Ti much faster than in pure titanium. In addition, it is assumed that the amount of nitrogen trapping by the surface of Fe-5Ti alloys does not depend on the immersion time.

For a strict control of tritium migration in fusion reactor fuel systems, tritium permeation barrier coatings have been developed for several decades. In liquid blanket concepts, corrosion of the coatings by liquid tritium breeders is a serious concern in addition to tritium permeation. Furthermore, the coatings are exposed to high-energy neutrons in an actual reactor, which would bring a synergy of irradiation and corrosion. In this study, zirconium oxide (ZrO2) coatings prepared by metal organic decomposition were exposed to static lithium-lead (Li-Pb) for 500 h at 500−600 oC after iron-ion irradiation with the damage concentration of up to 16 displacement per atom at room temperature and 500 oC. Corrosion products formed on all the exposed samples, and their amount and stability depended on the damage density and temperature during irradiation and Li-Pb exposure. Voids were formed by the irradiation and segregated in grain boundaries of the ZrO2 coating. The thicknesses of irradiated coatings decreased with damage concentration; therefore, Li-Pb corrosion of the coatings would be promoted by microstructure change caused by heavy-ion irradiation. These results suggest that the Li-Pb compatibility of the ZrO2 coatings would be degraded under irradiation environments.

Measuring methods of spatial neutron distribution have been required for the evaluation of TBR inside the blanket. In this paper, neutron transport inside a blanket mock-up, mainly assembled by beryllium neutron multiplying layers and lithium breeding layers, was simulated for measurements of the neutron energy spectrum and the spatial distribution with activation analyses using activation foils and an imaging plate. Transport of deuterium-tritium (DT) fusion reaction neutrons were simulated by MCNP6 and activation calculations were performed by DCHAIN. These results of neutron transport indicate that activation foils were reacted to neutrons with specific energies; dysprosium for thermal, manganese with cadmium for epi-thermal and niobium or cobalt for fast neutrons. The results of activation calculations indicate each activity of activation foils inside the blanket, which proposes the exposure conditions for measurements using imaging plate.

H2 permeation behavior in a new molten salt of Flibe mixed with 0.5 wt% Ti powder is experimentally and analytically investigated. Ti powder included in Flibe can suppress H2 permeation for...

A new liquid metal divertor system named the REVOLVER-D (Reactor-oriented Effectively VOLumetric VERtical Divertor) is proposed for the helical fusion reactor FFHR-d1. The REVOLVER-D is composed of molten tin shower jets stabilized by internal flow resistances of wire/tape/chain. These shower jets are inserted into the ergodic layer surrounding the main plasma. Tin is selected as the liquid metal because of its low melting point, low vapor pressure, low material cost, and high safety. The liquid metal pumps, cryopumps, and turbo molecular pumps are installed in the central vacuum vessel connected to the main vacuum vessel via 10 inner ports equipped with maze neutron shields. Central solenoid coils made of high-temperature superconductors are installed inside the central vacuum vessel to shield the pumps from the strong magnetic field. The REVOLVER-D has a good possibility to satisfy important characteristics required for the divertor system in a fusion reactor, that is, high heat load tolerance, high maintainability, sufficient vacuum pump speed, high level of safety, and a small amount of radioactive wastes.

The lead-lithium eutectic alloy (PbLi) is a promising coolant for nuclear fusion reactors. Although PbLi flow and heat transfer in a nuclear fusion reactor condition needs to be comprehended, no ve...

Our group is developing an ultra-compact neutron source based on inertial electrostatic confinement (IEC) fusion device for various applications at Kyoto University. This IEC device is configured from a titanium anode and a molybdenum cathode with diameters of 17 and 6 cm, respectively. A high-intensity neutron source operated in a stable pulse shape is mandatory to increase the system’s reliability. Applying a higher voltage is a straightforward way to increase the neutron yield from the system. However, a contradiction between the increase of the applied voltage and the reduction of the system size limits such a proposal. A three-stage feedthrough system is employed in the developed compact IEC to address this contradiction. A feedback control system was developed and applied to the input and output parameters, such as the applied voltage and the neutron yield, to increase its stability in long-term operation. Characterization of the developed system was performed by scanning the neutron yield as a function of applied voltage and cathode current. To date, a maximum neutron yield of 9.2 × 107n·s–1 at 6.4 kW (80 kV and 80 mA) has been obtained. A study of the feasibility of using the IEC system for neutron radiography was performed. Preliminary analysis of the resulting images showed there was good contrast between the sample and the background. The results suggest that optimization of the experimental parameters is needed to perform higher accuracy neutron radiography.

In order to establish a liquid blanket using a Li-containing molten fluoride salt, the corrosion rate and mechanism of pure Fe and JLF-1 steel in FLiNaK (LiF-NaF- KF = 46.5-11.5-42.0) was investigated by one of the electrochemical methods, the potential step method. Simulating the TF breeding in the blanket, HF gas was injected into FLiNaK at 773 K. The partial pressure of HF gas (pHF) was varied from 8.59 to 2.46 × 103 Pa. The current-potential curves of pure Fe and JLF-1 steel were measured in an electrochemical cell and fit by the Butler-Volmer equation with limited current density. The corrosion current density (icor) of pure Fe increased with pHF as i c o r = 0.6 p H F . The corrosion rates of pure Fe and JLF-1 steel were close. In order to keep the uniform corrosion rate of JLF-1 steel below 0.1 mm/year in the blanket, HF concentration in FLiNaK is required less than 1.2 × 10−4 molHF/m3. The cathodic current density (iLc) was proportional to pHF as i L c = - 0.11 p H F , and, therefore, the rate-limiting process of HF corrosion of pure Fe is predicted to be HF or H+ diffusion in FLiNaK.

In the lithium loop of IFMIF (International Fusion Materials Irradiation Facility), the target value of the nitrogen concentration in lithium flow is designed to be 99.9999 %), it was contaminated by nitrogen as an impurity of the cover gas. Therefore, the hot trap experimental pot sealed was placed in a vacuum exhaust glove box. In order to minimize nitrogen contamination, lithium sampling for nitrogen concentration measurement was carried out by scraping out lithium in a solid state. Then, nitrogen concentration was determined by the ammonia conversion method. As a result of the hot trap experiment under the above conditions, the nitrogen concentration in lithium was reduced from 43 to 10 wt.ppm by immersing Fe-5at.%Ti grains (∼160 μm) at 509 ℃ after 192 h, from 62 to 9.8 wt.ppm at 550 ℃ after 168 h, and from 65 to 9.9 wt.ppm at 557 ℃ after 144 h.

The hydrogen solubility in a FLiNaK molten salt mixed with Ti powder was investigated. A hydrogen-soluble metal powder mixed with a molten salt can increase the effective hydrogen solubility of the molten salt, which is currently a major disadvantage of molten salts. A significant increase in hydrogen solubility was observed, even with a mass fraction of Ti powder of only 0.1 wt%. The increase of hydrogen solubility was so large that a vanadium alloy would be compatible with the Ti powder/molten salt mixture, unlike typical molten salts that result in an unacceptably large tritium inventory in the vanadium alloy. In addition, contamination of the Ti powder by oxidation suppressed the hydrogen uptake and release capability. Controlling the purity of the molten salt was found to be one of the key issues for the metal powder mixture concept.

NIFS launched the Fusion Engineering Research Project (FERP) in preparation for DEMO by starting the redesign of the LHD-type helical reactor FFHR-d1. In the first round, the main parameters were selected. The second round is preparing detailed three-dimensional (3D) design of the superconducting magnet support structures, and 3D neutronics analyses, where the diverter targets can be efficiently shielded from fast neutrons. A new Flinabe blanket mixed with metal powder was proposed. Fabrication of helical coils by connecting half-helical-pitch segments of 100 kA-class YBCO high-temperature superconductors is proposed as a promising method. Also in progress is improvement of the first round of the core plasma design, ignition start-up analyses, and fueling scenario. As a consequence, a multi-path strategy on FFHR-d1 has been introduced with versions of -d1A, -d1B, and -d1C, where design flexibility is expanded to include subignition with options FFHR-c1 for “before demo, compact, and component-test.”

Hydrogen solubility in liquid Li–Na alloys (20, 50, and 90 at% Li) at 673, 773, and 873 K was investigated by adding and extracting hydrogen through the wall of a capsule made of pure iron. Hydrogen solubility in the Li90Na10 alloy was almost similar to that in pure Li while for Li50Na50 and Li20Na80, it was one and two orders of magnitude smaller, respectively. No obvious difference in temperature dependence was observed, similar to the case for pure Li. Li–Na alloy can thus be a potential liquid breeding material with almost all of the advantages of the Li blanket system and free of most of its drawbacks.

Release behaviors of tritium from molten FLiNaBe (LiF–NaF–BeF2) have been examined in a series of low flux neutron irradiation experiments with an AmBe neutron source at the OKTAVIAN facility of Osaka University. Tritium released from FLiNaBe is swept by He and Ar gases, and recovered by water bubblers. The recovery rate of tritium released from 300 cm3 of FLiNaBe was 58.0 mBq/h in a steady state, and it was close to the tritium production rate of 58.0 mBq/h calculated by the MCNP6 code. The values of the overall mass transfer coefficients for tritium release from FLiNaBe were close to those for FLiNaK (LiF–NaF–KF) at 773 K. The influence of the diffusivity and the solubility of tritium on the overall mass transfer coefficients has been examined on molten FLiNaBe.

Titanium plates coated by iron were fabricated electrochemically as a candidate hydrogen isotope permeation window for liquid lithium. Contacting the window to liquid lithium containing deuterium between 673 and 873 K, its permeation behavior and resistance to degradation were investigated. The iron-coated window showed less permeability than the bare titanium window as well as good agreement with the theoretical value. Furthermore, the iron layer at the vacuum side surface strongly improved oxidation resistance. Decrease of permeability for 2 hours air exposure could be recovered within 30 hours in the case of the coated window, while that of the bare titanium window could not. The coated permeation window is promising regarding high permeability and antidegradation properties.

Hydrogen permeation window is a candidate not only as the tritium recovery device from liquid lithium, but also as the hydrogen isotopes monitor for molten lithium because hydrogen permeation rate is dominated by hydrogen concentration. In this study, changing tritium concentration in lithium, tritium permeation through iron window contacting with tritium was quantified by ionization chambers. As a result, measurable amount of tritium permeated the window even though tritium concentration in lithium was in the order of wppb. The amount of tritium permeation increased as tritium concentration in lithium increased. Iron window is hopeful as tritium monitor for liquid lithium.

Sacrificial anode methods are focused upon for suppression of the tritium fluoride corrosion of structural materials in molten salt blankets in fusion reactors. Stainless Steel 430 (SS430) is immersed into hydrogen fluoride (HF)-containing LiF-NaF-KF (FLiNaK) for 6 hours at 773 K with Ti. It has been found that (1) the presence of Ti suppresses the corrosion of SS430 by HF. Further, by the electrical connection between SS430 and Ti, (2) Ti can sufficiently lower the potential of SS430, (3) the weight loss of Ti increases, and (4) the grain boundary corrosion of SS430 by Cr dissolution is mitigated. The applicability of sacrificial anode methods to molten salt blankets has been established.

As the finalization of a hydrogen experiment towards the deuterium phase, the exploration of the best performance of hydrogen plasma was intensively performed in the large helical device. High ion and electron temperatures, Ti and Te, of more than 6 keV were simultaneously achieved by superimposing high-power electron cyclotron resonance heating onneutral beam injection (NBI) heated plasma. Although flattening of the ion temperature profile in the core region was observed during the discharges, one could avoid degradation by increasing the electron density. Another key parameter to present plasma performance is an averaged beta value $\left\langle \beta \right\rangle $ . The high $\left\langle \beta \right\rangle $ regime around 4% was extended to an order of magnitude lower than the earlier collisional regime. Impurity behaviour in hydrogen discharges with NBI heating was also classified with a wide range of edge plasma parameters. The existence of a no impurity accumulation regime, where the high performance plasma is maintained with high power heating >10 MW, was identified. Wide parameter scan experiments suggest that the toroidal rotation and the turbulence are the candidates for expelling impurities from the core region.

EVEDA Lithium Test Loop (ELTL) has been designed and constructed, has operated a liquid lithium flow test facility with the world's highest flow rate and has succeeded in generating a 100-mm-wide and 25-mmthick free-surface lithium flow along a concave back plate steadily at a high speed of 20 m/s at 300°C for the first time in the world. This result will greatly advance the development of an accelerator-based neutron source to high energy and high density, one of the key objectives of the fusion reactor materials development under the BA (Broader Approach) Activities. Recent related engineering validation and engineering design of the lithium facility has been evaluated.

The hydrogen-release behavior of the molten salt FLiNaK containing directly synthesized 0.1 wt% nano-Ti powder was investigated. Hydrogen release was faster at higher temperatures, and the volume of released hydrogen from FLiNaK was significantly higher than that from pure FLiNaK. However, the ratio of released H/Ti in FLiNaK was smaller than that in μm-Ti powder.

Hydrogen Plasma-driven Permeation (PDP) experiments through two different liquid metal membranes: lithium and GaInSn have been conducted in the temperature range from 300 to 500 ◦C. A technique employing a mesh sheet to hold a liquid metal for PDP has been utilized for the first time. It has been found that PDP is surface recombination limited for lithium and is diffusion limited for GaInSn. Hydrogen surface recombination coefficients for liquid lithium and hydrogen diffusivity in GaInSn have been obtained respectively.

The subject of hydrogen isotopes transport through tungsten coated reduced activation ferritic steels such as F82H has attracted increasing interest in the fusion engineering research community. This paper reports on laboratory-scale studies that have been done to assess the hydrogen permeation properties of vacuum plasma-sprayed tungsten (VPS-W) coatings at the temperature range of 200 - 500 °C. W coatings with thicknesses of 46 μm and 90 μm have been investigated. It has been found that the observed permeation rates through composite VPS-W/F82H specimen are reduced to ∼7% compared to that of pure F82 H. VPS-W coating is porous and has an open system of connected pores, which density is evaluated to be ∼7%. The main effect of the W coating on hydrogen permeation is to reduce the incoming flux at the W/F82H interface owing to pore diffusion in the coating and to reduce the effective surface area for hydrogen dissolution in the substrate.

The lithium target facility of IFMIF (International Fusion Materials Irradiation Facility) consists of target assembly, lithium main loop, lithium purification loops, the diagnostic systems, and remote handling system. Major impurities in the lithium loop are proton, deuterium, tritium, 7-Be, activated corrosion products and the other species (C, N, O). It is very important to remove nitrogen content in lithium loop during operation, in order to avoid the corrosion/erosion of the nozzle of lithium target for the stable lithium flow on the target assembly. Nitrogen in the lithium can be removed by N hot trap using Fe-5at\%Ti alloy at temperatures from 400 to 600$^{\circ}$C. In this study, the specification and the detailed design were evaluated, and the component of N hot trap system was fabricated.