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Production of tritium using a high-temperature gas-cooled reactor (HTGR) has been studied for a prior engineering test with tritium handling and for the startup operation of a demonstration fusion reactor. For this purpose, the hydrogen absorption speed of Zr in a Li-loading rod for the reactor operation is experimentally measured, and an analysis model is presented to evaluate the tritium outflow from the Li rod in a high-temperature engineering test reactor (HTTR). On the basis of the presented model, the structure of the Li-loading rod for the demonstration test using the HTTR is proposed.

The neutron emission spectrum can be modified from Gaussian due to excited Alfven eigenmodes (AEs). We evaluate the neutron emission spectrum when the fuel-ion velocity distribution functions are distorted by AEs, assuming an ITER-like deuterium–tritium tokamak plasma. When shear AEs are excited by energetic particles in a plasma, the non-Maxwellian tails are created in the fuel-ion distribution functions in directions that are parallel and antiparallel to the magnetic field lines. Due to formation of these tails, the non-Gaussian components are formed in the neutron emission spectrum in the same and opposite directions of the toroidal axis. The anisotropic, non-Gaussian neutron emission spectrum can be used for experimental verification and diagnostics of the non-Maxwellian tails in the fuel-ion velocity distribution functions formed by AEs.

Analysis and experimental validation of energetic non-Maxwellian components formed in ion velocity distribution functions enhance our understanding of important physical phenomena in burning plasmas. As one such energetic component, knock-on tails appear due to nuclear elastic scattering (NES). It is important to experimentally validate the NES effects and analysis model of knock-on tails. For this purpose, we propose a method for determining the shape and size of a knock-on tail from the Doppler-broadened γ-ray emission spectrum in a proton-beam-injected deuterium plasma confined in an ITER-like experimental fusion device. In the proposed method, we adopt a two-temperature Maxwellian model and utilize the γ-ray-generating 6Li(d,nγ)7Be and 6Li(d,pγ)7Li nuclear reactions. We discuss both the validity of the proposed method and the expected accuracy of the shape and size of the measured knock-on tail.

It is well known that the influence of nuclear force, e.g., nuclear elastic scattering (NES), appears in an ion scattering process when ion energy increases. NES caused by high-energy particles forms the non-Maxwellian component in the ion distribution function. The non-Maxwellian component is named knock-on tail. The knock-on tail gives various effects on fusion plasma. Therefore, it is important to conduct an experiment in order to understand the impact of the NES effects. The NES effects are affected by the “shape and size” of the knock-on tail and the “shape and size” of the knock-on tail depends on plasma conditions. Therefore, we newly propose the method to observe the correlation between the NES effects and the plasma conditions by using Doppler effects for the γ-ray-generating reaction, i.e., 6 Li(d, p) 7 Li nuclear reaction, in ITER-like deuterium plasma. We can capture the impact of the NES effect caused by the change in the plasma conditions, i.e., electron temperature, deuteron density, proton beam power, and proton beam energy by capturing the change in the γ-ray spectrum. On the basis of the Boltzmann-Fokker-Planck model, we showed the validity of the method by simulation.

A high-temperature, gas-cooled reactor (HTGR) is proposed as a tritium production device that has the potential to produce a large amount of tritium using the 6Li(n,α)T reaction without major changes to the original reactor core design. In an HTGR design, generally, boron is loaded into the core as a burnable poison to suppress excess reactivity. In this study, lithium is loaded into the HTGR core aiming to produce thermal energy and tritium simultaneously and is loaded instead of boron as a burnable poison. The nuclear characteristics and fuel temperature were analyzed to confirm the nuclear and thermal feasibility of a lithium-loaded HTGR. It was shown that the analysis results satisfied the design requirements and hence the nuclear and thermal feasibility was confirmed for a lithium-loaded HTGR that produces thermal energy and tritium.

In high-temperature plasma sustained by neutral beam injection heating, the emission spectrum of the neutrons produced by fusion reactions is known to have a modified Gaussian distribution. In this study, neutron transport simulation is carried out for a plasma source that emits neutrons with a modified spectrum and the influence of this modified neutron emission spectrum on the tritium production performance is investigated. The results show that in tritium production using D(d,n)3He neutrons, the rate of the 9Be(n,2n)2α neutron multiple reactions, which has a threshold energy of 2.0 MeV, is significantly enhanced compared with that when a source with a Gaussian neutron emission spectrum was assumed. In addition, the influence of the blanket composition on the tritium production performance is discussed.

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 LiAlO2, 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.

To launch a fusion reactor stably, it is necessary to prepare sufficient amount of tritium by an external tritium source. Tritium production using nuclear reactions by neutron and Li in a high temperature gas-cooled reactor (HTGR) is an attractive method. An important issue is tritium confinement in high temperature conditions of HTGR. Covering Li compound by Al2O3 is a promising method because hydrogen permeability in Al2O3 is quite low. Furthermore, it is expected that inserting Zr between Li compound and Al2O3 suppresses tritium permeation because Zr has a large hydrogen absorption capacity. For example, a cylindrical multilayer structure consisted of Zr – Li compound – Zr – Al2O3 is a candidate structure to confine generated tritium. To evaluate hydrogen permeation rate through cylindrical Zr material, in this work, hydrogen permeation experiments were carried out for two samples of one-side sealed Zr pipe. Hydrogen permeation rate in one sample was proportional to as the square root of hydrogen pressure but that in another sample, which had retained a certain amount of hydrogen before the experiment, did not indicate an obvious pressure dependence regardless of almost the same experimental procedure except hydrogen permeation direction. Since observed hydrogen permeation rate in Zr was faster than that in Al2O3, under the multi confinement structure by Zr – Li compound – Zr – Al2O3, the generated tritium in Li compound diffuses promptly in whole Zr inside Al2O3 layer and it is expected tritium is stored stably in the structure. Keywords: Hydrogen, Permeation, Zirconium, Pipe

In the large helical device (LHD) deuterium plasma experiment, the neutron emission rate and the shot-integrated neutron yield are measured with the neutron flux monitor (NFM) and the neutron activation system (NAS), respectively, where the neutron emission is assumed to be isotropic in the plasma. The differential cross section of the D(d,n) 3 He reaction has a large anisotropy for the forward direction of the incident deuteron direction. LHD has intensive tangential neutral beam injectors (NBIs), which may cause an anisotropic neutron emission in the plasma. The angular distribution of the neutron emission is calculated from the fast-ion distribution function evaluated by a code that solves Fokker-Plank equations for the 180-keV tangential NBI. The effect of the anisotropic neutron emission on the NFM and NAS measurements is estimated by a general-purpose Monte Carlo N-particle (MCNP) code calculations. Also, the effect is confirmed experimentally. The neutron emission rate measured with the NFM near the equatorial port is about 10% larger than that with the NFM at the top of the LHD center axis. The shot-integrated neutron yield measured with NAS is 25% larger than that with NFM at the top of the LHD center axis in the case of the tangential neutral beam injection, which is consistent with the MCNP calculation.

A possible scenario to observe a knock-on tail using anisotropic neutron emission spectrum in a proton-beam injected deuterium plasma is presented. On the basis of the ion trajectory analysis in ITER-like magnetic configuration, Boltzmann collision integral and Fokker-Planck simulation, the knock-on tail formation in deuteron distribution function due to nuclear elastic scattering, and the resulting modification of the “double differential” neutron emission spectrum are examined. As a result of the beam injection with a specific direction, the neutron emission spectrum is modified in 2-D phase space. It is shown that the ratio of neutron emission rate with >2.8-MeV energy to the 2.45-MeV peak is increased almost 10 times compared with the 1-D modified neutron emission spectrum integrated over the emission angle, which implies the improvement of the accuracy in the neutron measurement. By using the anisotropic neutron emission adjusting the relative positions between beam-injection port and the neutron detector, the possibility to catch the knock-on tail for various plasma conditions can be increased.

The production of tritium (T) using high-temperature gas-cooled reactors (HTGRs) has been studied for a prior engineering research with T handling and initial T possession in demonstration fusion reactors. Stable containment of T in Li-loading rods during HTGR operation is a critical issue. This study investigates this for an irradiation test to examine T-containment performance in Li-loading rods and develops an analytical model of evaluating the amount of T outflow to a He coolant. The hydrogen absorption characteristics, including the deterioration of the hydrogen absorption speed after Zr has sufficiently absorbed the hydrogen, is experimentally measured assuming an HTGR setting. We present an analytical model of evaluating the T outflow from a Li rod and, on the basis of this model, estimate the total T outflow, assuming the presence of a gas-turbine high-temperature reactor of 300 MWe with a nominal capacity and a high-temperature engineering test reactor. It is demonstrated that, by loading a sufficient amount of Zr into the Li rod, the T outflow can be suppressed to less than a small percent of the total T produced during 360 days of reactor operation.

Performance of tritium production for fusion reactors, using a high-temperature gas-cooled reactor (HTGR) is examined. From the viewpoints of tritium recovery and environmental safety, tritium outflow from Li rods should be suppressed to the same level as the liquid radioactive waste from the pressurized water reactors (PWRs) in Japan. Methods for suppressing tritium leakage from Li rods are studied. The tritium outflow is reevaluated accurately on the basis of non-equilibrium simulations and the influence of coolant temperature on tritium leakage is clarified. The approach using Zr in the Li rod to reduce the tritium pressure and the resulting suppression of tritium leakage are also investigated.The results of the non-equilibrium simulation show that the tritium outflow is approximately 40% lower than the outflow reported in a previous study. Although the electric power generation efficiency is reduced, lowering the coolant temperature to 600 K results in a reduction of the tritium outflow to ~1/...

Nuclear reactors produce a great number of electron antineutrinos mainly from beta-decay chains of fission products (FP). Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may have information on fuel burn-up and may be detected in the future with advanced measurement technology. We calculate reactor antineutrino spectra especially in the low-energy region. In this work, we present neutrino spectra from various reactors such as typical pressurized water reactor and others types of reactors for comparison. Our result shows the electron antineutrino flux in the low-energy region increases with burn-up of nuclear fuel by accumulated nuclides with low Q values in beta-decay.

The fast deuteron (non-Maxwellian component) diagnostic method, which is based on the higher resolution optical spectroscopic measurement, has been developed as a powerful tool. Owing to a decrease in the D–H charge-exchange cross section, the diagnostic ability of conventional optical diagnostic methods should be improved for ∼MeV energy deuterons. Because the 3He–H charge-exchange cross section is much larger than that of D–H in the ∼MeV energy range, the visible light (VIS) spectrum of 3He produced by the dueteron–dueteron (DD) reaction may be a useful tool. Although the density of 3He is small because it is produced via the DD reaction, improvement of the emissivity of the VIS spectrum of 3He can be expected by using a high-energy beam. We evaluate the VIS spectrum of 3He for the cases when a fast deuteron tail is formed and not formed in the ITER-like beam injected deuterium plasma. Even when the beam energy is in the MeV energy range, a large change appears in the half width at half maximum of the VIS spectrum. The emissivity of the VIS spectrum of 3He and the emissivity of bremsstrahlung are compared, and the measurable VIS spectrum is obtained. It is shown that the VIS spectrum of 3He is a useful tool for the MeV beam deuteron tail diagnostics.

In this work, energetic-ion confinement and loss due to energetic-ion driven magnetohydrodynamic modes are studied using comprehensive neutron diagnostics and orbit-following numerical simulations for the Large Helical Device (LHD). The neutron flux monitor is employed in order to obtain global confinement of energetic ions and two installed vertical neutron cameras (VNCs) viewing different poloidal cross-sections are utilized in order to measure the radial profile of energetic ions. A strong helically-trapped energetic-ion-driven resistive interchange mode (EIC) excited in relatively low-density plasma terminated high-temperature state in LHD. Changes in the neutron emission profile due to the EIC excitation are clearly visualized by the VNCs. The reduction in the neutron signal for the helical ripple valley increases with EIC amplitude, which reaches approximately 50%. In addition to the EIC experiment, orbit-following simulations using the DELTA5D code with EIC fluctuations were performed to assess the energetic-ion transport and loss. Two-dimensional temporal evolution results show that the neutron emissivity at the helical ripple decreases significantly due to the EIC. The rapid reduction in neutron emissivity shows that the helically-trapped beam ions immediately escape from the plasma. The reduction in the VNC signals for the helical ripple valley and the total neutron emission rate increase with increasing EIC amplitude, as observed in the experiment. Calculated line-integrated neutron emission results show that the profile measured by VNC1 has one peak, whereas the profile measured by VNC2 has two peaks, as observed in the experiment. Although the neutron emission profile for VNC2 has a relatively wide peak compared with the experimental results, the significant decrease in neutron signal corresponding to the helical ripple valley was successfully reproduced.

Tritium production utilizing nuclear reactions by neutron and lithium in a high-temperature gas-cooled reactor is attractive for development of a fusion reactor. From viewpoints of tritium safety a...

To start up a fusion reactor, it is necessary to provide a sufficient amount of tritium from an external device. The fusion DEMO reactor is planned to start up in the 2030s. Herein, methods for supplying the reactor with tritium are discussed. For the initial startup of the fusion reactor, use of a high temperature gas-cooled reactor (HTGR) as a tritium production device has been proposed. So far, the analyses have been focused only on the operation in which fuel is exchanged at stated periods (batch) using the block type HTGR. In this paper, to improve the performance of tritium production, properties of the HTGR are studied from the viewpoint of continuous operation for several conditions. In continuous operation, for example, in the pebble bed type HTGR, it is possible to design an operation that has no time loss for refueling. The pebble bed modular reactor (PBMR) and the gas turbine high temperature reactor of 300 MWe nominal capacity (GTHTR300) are assumed as the calculation and comparison targets, ...

The performance of a high-temperature gas-cooled reactor as a tritium production device was examined. A gas turbine high-temperature reactor of 300 MWe nominal capacity (GTHTR300) was assumed as th...

A scenario to verify the knock-on tail formation by energetic protons and the resulting enhancement of the fractional ion heating by using 3He-containing deuterium plasma is presented. On the basis of the Boltzmann-Fokker-Planck (BFP) model, the knock-on tail formation in deuteron distribution function due to nuclear plus interference (NI) scattering of fusion-produced energetic protons and the resulting modification of the neutron emission spectrum are examined. A recognizable change in the energetic-neutron emission rate due to the knock-on tail formation depending on the external 3He inclusion is shown. Another possible scenario to ascertain the enhancement of ion-heating by using γ-ray-generating 6Li+D reaction is also presented and discussed.

Fuel-ion velocity distribution functions can be distorted from a Maxwellian by Alfven eigenmodes when the mode amplitudes are large, and the fusion reaction rate coefficient can be enhanced by the distortion. Assuming an ITER-like deuterium–tritium plasma containing an excited Alfven eigenmode, we have investigated the modified deuteron and triton velocity distribution functions and the resulting fusion reaction rate coefficient. It is found that the non-Maxwellian components in the distribution functions form in directions parallel and antiparallel to the magnetic field lines due to the toroidal Alfven eigenmodes or the ellipticity-induced Alfven eigenmodes. Focusing on the effect of the distortion of the distribution functions, we discuss the dependence of the enhancement of the fusion reaction rate coefficient on the mode amplitudes and determine the threshold amplitude necessary to enhance the rate coefficient.

A new method to diagnose the degree of electron degeneracy in compressed fuel for fast-ignition inertial confinement fusion is proposed We focus on 4.44-MeV γ-rays emitted in the reaction 9 Be(a, nγ) 12 C governed by fusion-produced energetic alpha-particles in a laser-imploded DT fuel pellet admixed with a small amount of 9 B. In this case the compressed fuel pellet is not subjected to any heating laser pulse. We have evaluated the probability P α-Be that the a+9B e reaction occurs during the slowing down of a-particle. It is found that the reaction probability depends strongly on the degeneracy parameter Θ, which is defined as the ratio of electron temperature to the Fermi energy. We show the possibility of diagnosing the electron degeneracy from the P α-Be - Θ diagram by detecting the 4.44-MeV γ-quanta and DT neutrons emitted from the dense core plasma.

An effect of the nuclear elastic scattering (NES) on the neutral beam injection (NBI) plasma heating was examined by solving the Boltzmann-Fokker-Planck (BFP) equation for beam ion in the deuterium-tritium (DT) thermonuclear plasmas. The BFP calculations show that the enhancement in the fraction of the NBI heating power deposited to ions due to NES becomes appreciable when beam energy is larger than 1MeV, and the enhancement is strongly influenced by plasma parameters.

A diagnostic method for small non-Maxwellian tails in fuel-ion velocity distribution functions is proposed; this method uses the anisotropy of neutron emissions, and it is based on the numerical analysis of the incident fast neutron spectrum on the first wall of a fusion device. Neutron energy spectra are investigated for each incident position along the first wall and each angle of incidence assuming an ITER-like deuterium–tritium plasma; it is heated by tangential-neutral-beam injection. Evaluating the incident neutron spectra at all wall positions and angles of incidence enables the selective measurement of non-Gaussian components in the neutron emission spectrum for energetic ion diagnostics; in addition, the optimal detector position and orientation can be determined. At the optimal detector position and orientation, the ratio of non-Gaussian components to the Gaussian peak can be two orders of magnitude greater than the ratio in the neutron emission spectrum. This result can improve the accuracy of energetic ion diagnostics in plasmas when small, anisotropic non-Maxwellian tails are formed in fuel ion velocity distribution functions. We focus on the non-Gaussian components greater than 14 MeV, where the effect of the background noise (i.e., slowing-down neutrons by scattering throughout the machine structure) can be ignored.A diagnostic method for small non-Maxwellian tails in fuel-ion velocity distribution functions is proposed; this method uses the anisotropy of neutron emissions, and it is based on the numerical analysis of the incident fast neutron spectrum on the first wall of a fusion device. Neutron energy spectra are investigated for each incident position along the first wall and each angle of incidence assuming an ITER-like deuterium–tritium plasma; it is heated by tangential-neutral-beam injection. Evaluating the incident neutron spectra at all wall positions and angles of incidence enables the selective measurement of non-Gaussian components in the neutron emission spectrum for energetic ion diagnostics; in addition, the optimal detector position and orientation can be determined. At the optimal detector position and orientation, the ratio of non-Gaussian components to the Gaussian peak can be two orders of magnitude greater than the ratio in the neutron emission spectrum. This result can improve the accuracy of ...

The radial profile of the neutron production rate in spherical inertial electrostatic confinement (SIEC) plasmas is numerically investigated for various device parameters, i.e., grid cathode current, grid voltage, etc. The electrostatic potential is obtained by solving the Poisson equation; and using the potential, the fuel-ion velocity distribution function is determined at each radial point. From the space-dependent velocity distribution function, the radial profile of the neutron production rate is evaluated. The influence of the broadness of the electron angular momentum distribution on the radial profile of the neutron production rate is also examined. It is shown that the height of the peak of the neutron production rate and its radial position are strongly influenced by the device parameters and the electron distribution.

The effect of nuclear elastic scattering (NES) on energetic alpha-particles two-dimensional (2D) velocity distribution function in a magnetic fusion device is examined by particle orbit simulation. Energetic alpha-particles exhibit an anisotropic 2D velocity distribution function, because the alpha particle loss reduces the amount of particles moving in a particular direction, and an alpha particle orbit depends on the initial direction of particle. When NES changes the direction of fast alpha particle motion, its orbit can be changed. It is shown that NES causes particle orbit transitions and changes the energetic alpha particles 2D velocity distribution function. We discuss the manner in which the distribution function is changed and how it affects the confinement of fast alpha particles.

Formation of an effective ion tail due to neutral beam injection heating during startup in D-[sup 3]He plasmas is investigated. The main idea is to reduce the energy input required for startup heating as well as the 14-MeV neutron yield by creating an effective tail. The optimal beam injection energy and beam species are first estimated by solving the steady-state Fokker-Planck equations for the injected species and for tritons. The startup of D-[sup 3]He plasma is simulated by simultaneously solving the time-dependent power balance and particle conservation equations together with the Fokker-Planck equations. As a result of tail formation in the fuel ion distribution, both the total input energy and the 14-MeV neutron yield during the startup phase are reduced by [approximately]20% from the values for Maxwellian plasma. 12 refs., 10 figs., 3 tabs.

Under tritium production method using a high-temperature gas-cooled reactor loaded Li compound, Li compound has to be coated by ceramic materials in order to suppress the spreading of tritium to the whole reactor. Pyrolytic carbon (PyC) is a candidate of the coating material because of its high resistance for gas permeation. In this study, hydrogen permeation experiments using a PyC-coated isotropic graphite tube were conducted and hydrogen diffusivity, solubility and permeability were evaluated. Tritium permeation behavior through PyC-coated Li compound particles was simulated by using obtained data. Hydrogen permeation flux through PyC in a steady state is proportional to the hydrogen pressure and is larger than that through Al 2 O 3 which is also candidate coating material. However, total tritium leak within the supposed reactor operation period through the PyC-coated Li compound particles is lower than that through the Al 2 O 3 -coated ones because the hydrogen absorption capacity in PyC is considerably larger than that in Al 2 O 3 .