Activation and transmutation of tungsten boride shields in a spherical tokamak

The FISPACT-II code is used to compute the levels of activation and transmutation of tungsten borides for shielding the central High Temperature Superconductor (HTS) core of a spherical tokamak fusion power plant during operations at 200 MW fusion power for 30 years and after shutting down for 10 years. The materials considered were W2B, WB, W2B5 and WB4 along with a sintered borocarbide B0.329C0.074Cr0.024Fe0.274W0.299, monolithic W and WC. Calculations were made within shields composed of each material, for five reactor major radii from 1400 to 2200 mm, and for six 10B isotope concentrations and at five positions across the shield. The isotopic production and decay in each shield is detailed. The activation of boride materials is lower than for either W or WC and is lowest of all for W2B5. While isotopes from tungsten largely decay within 3 years of shut-down, those from boron have a much longer decay life. An acceptable 70% of the absorbing 10B isotope will remain after 30 years of operations behind the first wall for a 1400 mm radius tokamak. Gaseous production is problematic in boride shields, where 4He in particular is produced in quantities 3 orders of magnitude higher than in W or WC shields. The FISPACT-II displacements per atom (dpa) tend to increase with boron content, although they decrease with increased 10B isotopic content. The dpa ranges of boride shields tend to lie between those of W and WC. Overall, the results confirm that the favourable fusion reaction shielding properties of W2B5 are not seriously challenged by its irradiation and transmutation properties, although helium gas production could be a challenge to its thermal and mechanical properties.

Overview of JET results for optimising ITER operation

The JET 2019-2020 scientific and technological programme exploited the results of years of concerted scientific and engineering work, including the ITER-like wall (ILW: Be wall and W divertor) installed in 2010, improved diagnostic capabilities now fully available, a major Neutral Beam Injection (NBI) upgrade providing record power in 2019-2020, and tested the technical & procedural preparation for safe operation with tritium. Research along three complementary axes yielded a wealth of new results. Firstly, the JET plasma programme delivered scenarios suitable for high fusion power and alpha particle physics in the coming D-T campaign (DTE2), with record sustained neutron rates, as well as plasmas for clarifying the impact of isotope mass on plasma core, edge and plasma-wall interactions, and for ITER pre-fusion power operation. The efficacy of the newly installed Shattered Pellet Injector for mitigating disruption forces and runaway electrons was demonstrated. Secondly, research on the consequences of long-term exposure to JET-ILW plasma was completed, with emphasis on wall damage and fuel retention, and with analyses of wall materials and dust particles that will help validate assumptions and codes for design & operation of ITER and DEMO. Thirdly, the nuclear technology programme aiming to deliver maximum technological return from operations in D, T and D-T benefited from the highest D-D neutron yield in years, securing results for validating radiation transport and activation codes, and nuclear data for ITER.

On the plasma quasineutrality under oscillatory confinement based on a nanosecond vacuum discharge

One of the main problems for inertial electrostatic confinement devices with electron injection is the space charge neutralization. This work is devoted to the analysis of the problem of plasma quasineutrality in the scheme of plasma oscillatory confinement based on nanosecond vacuum discharge (NVD). Electrodynamics modeling of the processes of aneutronic fusion of proton–boron showed that the plasma in the NVD, and especially on the discharge axis, really corresponds to a quasineutral regime, which is rather different from the well-known scheme of periodically oscillating plasma spheres (POPS). In this case, small oscillations in the NVD are a mechanism of resonant ion heating, unlike coherent compressions in the original POPS model. The scaling of the fusion power turns out to be close to the fusion scheme with POPS, but differs significantly in the values of the parameter of quasineutrality and the compression ratio.

Electric Power Amplification in Fusion Power Plants

Fusion power concepts that are heated by electrical devices for the purpose of producing high levels of electrical output are in effect electric power amplifiers. Three systems are considered: A hypothetical electric power version of the ITER experiment, the ARIES-1 fusion reactor design, and a modified version of ARIES-1 with stainless steel structural material. We find that an ITER power plant with a reasonable electric power conversion system would produce no net electric power at its target energy amplification factor of 10. The ARIES-1 conceptual power plant, as conceived, would have an energy amplification of 22 and an electric amplification of 6. If stainless steel were substituted for the SiC composite material assumed, the ARIES-1 electric power amplification would drop to roughly 3. We conclude that practical fusion power plants will likely require a near-ignition operating mode and qualified high temperature materials as prerequisites for commercial viability.

Fusion firms target electricity generation by the 2030s

Most private fusion companies expect fusion power to be supplying electricity to the grid in the 2030s, according to an industry report.

Prediction of divertor heat flux width for ITER Pre-Fusion Power Operation using BOUT++ transport code

Prediction of divertor heat flux width is performed for the first and the second Pre-Fusion Power Operation (PFPO) phases specified in the new ITER Research Plan using BOUT++ transport code [Li N.M. et al 2018 Comput. Phys. Commun. 228 69–82]. The initial plasma profiles inside the separatrix are taken from CORSICA scenario studies. Transport coefficients in transport code are calculated by inverting the plasma profiles inside the separatrix and are assumed to be constants in the scrape-off-layer (SOL). An anomalous thermal diffusivity scan is performed with E×B and magnetic drifts. The results in two scenarios identify two distinct regimes: a drift dominant regime when diffusivity is smaller than the respective critical diffusivity χc and a turbulence dominant regime when diffusivity is larger than it. The Goldston heuristic drift model and the ITPA multi-machine experimental scaling yield a lower limit of the width λq. From transport simulations, we obtain the critical diffusivity χc = 0.5 m2⁄ s in 5MA/1.77T PFPO-1 scenario and χc = 0.3 m2⁄ s in 7.5MA/2.65T PFPO-2 scenario. Separatrix temperature and collisionality also have a significant impact on the heat flux width in the drift dominant regime. The investigation clearly yields a scaling for critical thermal diffusivity χc ∝ A½ ⁄ ((Z(1+Z)½ Bp 2)) using ITER scenarios with fixed safety factor q95, major radius R, aspect ratio R/a, and the separatrix temperature T, as well as established the connection with CFETR and C-Mod discharges. This scaling implies that for a given tokamak device with q95, R, R/a, and T fixed, a reduction of poloidal magnetic field by a factor of 3 leads to a 9 times higher critical value of thermal diffusivity χc, possibly yielding a transition from turbulence to drift dominant regime.

Scaling of DD fusion power in a nanosecond vacuum discharge

Earlier, in a nanosecond vacuum discharge (NVD) with a deuterated Pd anode, the appearance of DD neutrons was observed not only at the well-studied quasi-stationary stage, where a virtual cathode (VC) appears in the interelectrode space, but also at the very initial stage of the discharge. An analysis of the experiment shows that the autoelectron beam can play the role of a kind of trigger for starting DD syntheses processes on the surface or in the bulk of the Pd anode, but its mechanism at the initial stage of the discharge remained unclear. In this work, we performed PiC modeling of the possible partial penetration of a beam of autoelectrons into hollow anode Pd tubes. This leads to the formation of very small short-lived VCs inside individual Pd tubes, where, starting from a current of 100 A, DD microsynthesis is possible. It is shown that in devices with oscillating ions the favorable scaling of the DD fusion power, which increases with decreasing VC radius, can be retained up to rVC  0.02 cm.

Research on Trusted Interconnection of Power Internet of Things Perception Terminal

Based on the perception technology of power Internet of Things, this paper selects software such as Load Runner 10.0 as the research tool, simulates the application of network software of 60 clients through four computers simultaneously, and detects its relevant running status. Studies have found that the majority of users can normal login, using the system operation, the complete set of business processes, but at the same time, this study also continue to explore complex business operation in the process of software development problems, on the basis of monophyletic Overlay the development road of processing nodes to join, the goal is to fusion power iota perception technology, promote the further development and application of iota perception software.