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The deuterium (D) permeation behavior for 1 displacement per atom Fe2+ damaged tungsten (W) was studied by the gas-driven permeation method and compared with undamaged W. The results of thermal des...

Geant4 Monte Carlo simulations were carried out to investigate the possible shielding materials of aluminum, polyethylene, hydrides, complex hydrides and composite materials for radiation protection in spacecraft by considering two physical parameters, stopping power and fragmentation cross section. The dose reduction with shielding materials was investigated for Fe ions with energies of 500 MeV/n, 1 GeV/n and 2 GeV/n which are around the peak of the GCR energy spectrum. Fe ions easily stop in materials such as polyethylene and hydrides as opposed to materials such as aluminum and complex hydrides including high Z metals with contain little or no hydrogen. Attenuation of the primary particles in the shielding and fragmentation into more lightly charged and therefore more penetrating secondary particles are competing factors: attenuation acts to reduce the dose behind shielding while fragmentation increases it. Among hydrogenous materials, 6Li10BH4 was one of the more effective shielding materials as a function of mass providing a 20% greater dose reduction compared to polyethylene. Composite materials such as carbon fiber reinforced plastic and SiC composite plastic offer 1.9 times the dose reduction compared to aluminum as well as high mechanical strength. Composite materials have been found to be promising for spacecraft shielding, where both mass and volume are constrained.

Water-cooled pebble-bed (WCPB) blanket, in which beryllium/beryllide in a pebble form is used as neutron multiplier, is one of blanket concepts based on conventional or near-future technology for fusion DEMO. Combination of water, as coolant, and beryllium/beryllide, however, may pose a critical safety problem, i.e. the chemical reactivity of the beryllium/beryllide pebble and hydrogen generation. We present a new phenomenological model of the reaction behavior of the beryllium/beryllide pebble with the steam. The model consists of the equations of the transients of (i) the radius of the unreacted part and (ii) the temperature of the pebble. We have developed a code PSYCHE to numerically solve the model equations. It has been found that the amount of the reaction-produced hydrogen obtained by the numerical simulation agree well with the experimental observations. We also show an application of the code to safety analysis of the transient behaviors of the Be and beryllide Be12Ti pebbles in an in-box LOCA, i.e. loss-of-coolant accident in a blanket box. The model simulation presents the better thermal stability of the Be12Ti pebble, compared to the Be pebble, in the in-box LOCA condition expected in a WCPB DEMO blanket.

To study the hydrogen isotopes plasma driven permeation (PDP) behavior in plasma facing materials, a linear Radio Frequency (RF) plasma device has been constructed in the radiation controlled area at Shizuoka University. The deuterium (D) plasma is generated by injecting RF power with the frequency of 13.56 MHz through a copper antenna and confined by DC magnetic field. The sample is sealed by gold (Au) coated O-ring and one side (upstream side) of sample is exposed to the D plasma. The other side of sample, named as downstream side, is pumped out by a turbo molecular pump and a rotary pump. The permeated D through the sample is monitored by a quadrupole mass spectrometer (QMS) which is connected to the downstream chamber. Infrared heater is adopted to control the sample temperature. The PDP experiments under different plasma parameters show that the permeation process agrees with RD regime. The D recombination coefficient on upstream surface of W is obtained.

Hydrogen addition to a sweep gas of a solid breeder blanket module has been proposed to enhance tritium recovery from the surface of the breeders. However, the influence of hydrogen addition on the bred tritium behavior is not understood completely. Tritium behavior in the simplified blanket module of Li 2 TiO 3 pebbles was numerically calculated considering diffusion in the grain bulk, surface reactions on the grain surface and permeation through the cooling pipe. Although a partial pressure of T 2 increases with increasing a partial pressure of H 2 in the sweep gas, it was estimated that tritium permeation rate to the cooling water decreases. Additionally, the release duration of water vapor generated by the reaction of the pebbles and hydrogen is shortened with increasing a partial pressure of H 2 . Tritium inventory in the grain bulk and the grain surface occupies 99.6 % of total tritium inventory in the blanket module.

Deuterium (D) plasma driven permeation (PDP) experiments for tungsten (W) samples were conducted by a linear radio frequency (RF) plasma device. In the PDP experiment, the W sample surface is perpendicular to the grain elongation direction. The D ion flux is in the order of 1021 m−2 s−1 determined by a double Langmuir probe. The results show that bias had a limited effect on the D plasma driven permeation behavior. The D recombination coefficient on W surface is obtained at the temperature ranging from 740 K to 1031 K. The experimentally measured recombination coefficient for a pristine W surface is lower than that for a clean W surface. The effect of recombination coefficient on the D permeation and retention behaviors in W are studied by Hydrogen Isotope Diffusion and Trapping (HIDT) simulation code. The low recombination coefficient leads to a high D concentration in W and a high permeation rate at the back surface. The D permeation rate through a 8 mm thick W with a pristine surface is estimated to be 8.1 × 1018 D m s−1 under the incident ion flux of 1 × 1024 m−2 s−1 and temperature of 1173 K.

We examine the effect of neutron irradiation on the release of deuterium from tungsten at 573 K to understand the efficiency of tritium removal by baking out at moderate temperatures. Tungsten samples, undamaged and neutron-irradiated to a damage level of approximately 0.016 displacements per atom, are exposed to low-energy (108 eV), high-flux (3.0 × 1021 to 9.4 × 1021 m−2 s−1) deuterium plasma at temperatures ranging from 573 to 773 K to an ion fluence of 1.1 × 1025 m−2. At each exposure temperature, two undamaged and two neutron-irradiated tungsten samples are exposed to plasma. The deuterium content in the tungsten samples is measured by thermal desorption spectrometry soon after the plasma exposure and after post-plasma annealing at 573 K for 30 h. It is found that: (i) the deuterium retention in the neutron-irradiated tungsten samples is significantly higher than that in the undamaged tungsten samples; (ii) annealing at 573 K of undamaged tungsten samples pre-exposed to deuterium plasma at 573–773 K leads to an almost complete (60%–99%) release of deuterium from the samples; (iii) annealing at 573 K of neutron-irradiated tungsten samples pre-exposed to deuterium plasma at 573–773 K leads to a significant (8%–20%) release of deuterium from the samples.

Vertical instability of an elongated plasma and its stabilized concepts by in-vessel components and vacuum vessel (VV) design have been studied intensely in JA and EU Broader Approach, DEMO Design Activity. The vertical stabilization of the plasma represents one of the key issues for EU and JA DEMO, due to the large distance of the active control coils for the presence of thick breeding blanket system. A feasible DEMO reactor that maintains plasma vertical stability was proposed from an engineering viewpoint. The vertical stability performances are acceptable, without considering a passive stabilizer, if a maximum elongation of κ95 = 1.6 is chosen. For the higher-elongated plasmas (κ95 > 1.70), additional inboard passive stabilizer is effective.

Reduced activation ferritic/martensitic (RAFM) steels have been developed as the structural material for the fusion demonstration reactor, DEMO. These steels contain elements that produce radioactive isotopes and decay to low levels in timeframe required by the waste management scenario. Developments within the past quarter-century suggest a practical limit to the removal of undesired impurities such as Co, Cu, Ni, Mo and Nb. The concentrations of elements essential for the mechanical properties of RAFM steels, such as Al and N, required a compromise between the waste disposal scenario and performance demand. The limits and potential of RAFM steel pertaining to reducing the activation levels after service are discussed based on the actual achievements of F82H, Japanese RAFM steel, and numerical analyses of the activity. It was found that in order to achieve the shallow land burial limits 100 years after a reactor shutdown, Ni is the most significant impurity that must be removed (Mo in the case of the first wall). Limiting N below concentrations of 100 ppm will not be possible for a large scale melt, but concentrations of Al up to the maximum amount that has been achieved present no problems.

For the water-cooled solid blanket of DEMO, the nuclear analysis was performed based on present cooling piping system. Especially, distributions of neutron load and temperature were calculated with Pn is 5 MW/m 2 . Furthermore, the local TBR was optimized by changing the material proportion for each Pn level (1–5 MW/m2). It was confirmed that the size of cooling loop for sub-critical water could be used as about 2000 × 450 mm and the cooling pipe diameter of D is 12 mm, d is 9 mm at v is 5.36 m/s. The pipe pitches would vary with Pn level which is related to the blanket structure design. Nuclear heat distribution is the base to decide the distribution of cooling pipe positions. It was found that the local TBR of blanket would be dropped down along with the Pn level rising which was mainly depended on the thickness of beryllium variation. Finally, the layout of cooling pipes for each level was obtained.

a b s t r a c t Nuclear analysis results were compared for water-cooled blanket based on PWR (pressurized water reactor) and SCWR (sub-critical water reactor) water conditions. The local TBR (tritium breeding ratio) in outboard zone was discussed in the range of Pn (neutron wall load) from 1 MW/m 2 to 5 MW/m 2 . It was found that water fraction has little impact on TBR, which is an important factor related to blanket tritium efficiency. It indicated that TBR value of each Pn would be similar under the two kinds of water conditions, but PWR case is a little higher than that of SCWR’s. In addition, it was found that beryllium is the dominant factor leading a higher TBR inside blanket. As a result, TBR is an insensitive value with the water condition variation. The results would be important to water condition choice for solid blanket in the future.

Power exhaust to the divertor and the conceptual design have been investigated for a steady-state DEMO in Japan with 1.5 GW-level fusion power and the major radius of 8.5 m, where the plasma parameters were revised appropriate for the impurity seeding scenario. A system code survey for the Ar impurity seeding suggested the volume-averaged density, impurity concentration and exhaust power from the main plasma of = 205–285 MW. The divertor plasma simulation (SONIC) was performed in the divertor leg length of 1.6 m with the fixed exhaust power to the edge of = 250 MW and the total radiation fraction at the edge, SOL and divertor ( = 0.8), as a first step to investigate appropriate design of the divertor size and geometry. At the outer target, partial detachment was produced near the strike-point, and the peak heat load () at the attached region was reduced to ~5 MW m−2 with appropriate fuel and impurity puff rates. At the inner divertor target, full detachment of ion flux was produced and the peak was less than 10 MW m−2 mostly due to the surface-recombination. These results showed a power exhaust scenario and the divertor design concept. An integrated design of the water-cooling heat sink for the long leg divertor was proposed. Cu-ally (CuCrZr) cooling pipe was applicable as the heat sink to handle the high heat flux near the strike-point, where displacements per atom rate was estimated to be 0.5–1.5 per year by neutronics calculation. An arrangement of the coolant rooting for Cu-alloy and Reduced Activation Ferritic Martensitic (RAFM) steel (F82H) pipes in a divertor cassette was investigated, and the heat transport analysis of the W-monoblock and Cu-alloy pipe under the peak of 10 MWm−2 and nuclear heating was performed. The maximum temperatures on the W-surface and Cu-alloy pipe were 1021 and 331 °C. Heat flux of 16 MW m−2 was distributed in the major part of the coolant pipe. These results were acceptable for the plasma facing and structural materials.

A new conceptual design of the laser fusion power plant FALCON-D (Fast-ignition Advanced Laser fusion reactor CONcept with a Dry wall chamber) has been proposed. The fast-ignition method can achieve sufficient fusion gain for a commercial operation (∼100) with about 10 times smaller fusion yield than the conventional central ignition method. FALCON-D makes full use of this property and aims at designing with a compact dry wall chamber (5–6 m radius). 1D/2D simulations by hydrodynamic codes showed a possibility of achieving sufficient gain with a laser energy of 400 kJ, i.e. a 40 MJ target yield. The design feasibility of the compact dry wall chamber and the solid breeder blanket system was shown through thermomechanical analysis of the dry wall and neutronics analysis of the blanket system. Moderate electric output (∼400 MWe) can be achieved with a high repetition (30 Hz) laser. This dry wall reactor concept not only reduces several difficulties associated with a liquid wall system but also enables a simple cask maintenance method for the replacement of the blanket system, which can shorten the maintenance period. The basic idea of the maintenance method for the final optics system has also been proposed. Some critical R&D issues required for this design are also discussed.

The fast ignition is quite attractive for a compact laser fusion reactor, because a sufficiently high pellet gain is available with a small input energy. We designed an inertial fusion reactor based on Fast-ignition Advanced Laser fusion reactor CONcept, called FALCON-D, where a dry wall is employed for a chamber wall. A simple point model shows that the pellet gain G~100 is available with laser energies of 350kJ for implosion, 50kJ for heating. This results in the fusion yield of 40 MJ in one shot. By increasing the repetition rate up to 30 Hz, the fusion power of 1.2 GWth becomes available. Plant system analysis shows the net electric power to be about 0.4 GWe In the fast ignition it is available to employ a low aspect ratio pellet, which is favorable for the stability during the implosion phase. Here the pellet aspect ratio is reduced to be 2 ~ 4, and the optimization of the pulse shape for the implosion laser are carried out by using the 1-D hydrodynamic simulation code ILESTA-1D. A ferritic steel with a tungsten armour is employed for the chamber wall. The feasibility of this dry wall concept is studied from various engineering aspects such as surface melting, physical and chemical sputtering, blistering and exfoliation by helium retention, and thermo-mechanical fatigue, and it is found that blistering and exfoliation due to the helium retention and fatigue failure due to cyclic thermal load are major concerns. The cost analysis shows that the construction cost is moderate but the cost of electricity is slightly expensive.

Neutron scattering and Mossbauer absorption techniques are applied to the mixture Fe x Ni 1− x Cl 2 in which the strong Ising-type anisotropy of Fe and the weak XY-type anisotropy of Ni complete with each other. A mixed ordering phase is found to exist in a narrow range of concentrations around x ≈ 0.08. It is found that if a small amount of Fe, ≈2% replaces Ni in NiCl 2 the directions of majority spins of Ni rise up from the c -plane (the spin easy axis of NiCl 2 ) as a whole Diffuse scattering with concentrations is observed in the S ‖ -ordered antiferromagnetic (AF) phase over a wide range of temperature for the samples with concentrations near the crossing point. It might be one of the distinctive properties of the ordered phase in the mixture with competing spin anisotropies.

Mossbauer studies show that the spin axis in RbFeF3 is along [111], [100] and [111] in the three magnetic phases, which appear at TN>T>T2, T2>T>T1 and TT>T1 is found to originate in a ferrimagnetic order of the two inequivalent Fe2+ sites. For the weak ferromagnetism along [110] observed in the interval T