An electrochemical hydrogen pump (EHP) with a proton exchange membrane (PEM) used as part of fusion cycle systems successfully combines the processes of hydrogen extraction, purification and compression in a single device. This work comprises a novel study of the effect of ionizing radiation on the properties of the PEM as part of the EHP. Radiation exposure leads to nonspecific degradation of membranes, changes in their structure, and destruction of side and matrix chains. The findings from this work reveal that the replacement of sulfate groups in the membrane structure with carboxyl and hydrophilic groups leads to a decrease in conductivity from 0.115 to 0.103 S cm⁻¹, which is reflected in halving the device performance at a temperature of 30 °C. The shift of the ionomer peak of small-angle X-ray scattering curves from 3.1 to 4.4 nm and the absence of changes in the water uptake suggested structural changes in the PEM after the irradiation. Increasing the EHP operating temperature minimized the effect of membrane irradiation on the pump performance, but enhanced membrane drying at low pressure and 50 °C, which caused a current density drop from 0.52 to 0.32 A·cm⁻² at 0.5 V.

A generalized analytical model for the turn-off process of bipolar switches with microgates is developed taking into account the role of technological and design imperfections of real structures, which limit their safe operating area. Using a thyristor microchip with external field control, operating in a voltage inverter circuit, as an example, the boundary of the safe operating area with respect to the “turned-off current” is quantitatively determined. It is found that this boundary for a non-ideal structure in the low-voltage region is controlled by the effect of regenerative unlocking [triggering] of the cathode emitter; in the high-voltage region, this boundary is controlled by the onset of current localization in “perturbed” cells, involving dynamic breakdown. The possible applications of the developed model to indicate the directions of device-structure optimization to increase their maximum switching current are discussed. The adequacy of the results of model applications is validated using numerical simulation.

The dependence of the neutron yield of the FNS-ST (spherical tokamak) fusion neutron source on the fraction of tritium in the core D+T plasma is analyzed for the case of using tritium neutral beam injectors with 200-keV energy and 6-MW power. The FNS-ST operating regimes are explored using the SOLPS4.3 and ASTRA codes for different values of core plasma density n_e, T fraction in the plasma, and particle diffusivity. The FC-FNS code is used to estimate the fluxes of the fuel components in the fuel cycle (FC), which are produced by different injection systems: gas puffing, pellet injection, and neutral beam (T) injection. It is shown that in the case of the Т beam injection, in the operating range of parameters, the neutron yield can reach 6.0 × 10¹⁷ s⁻¹, which is the value comparable to that obtained for the scenario of D-beam injection into the balanced D+T plasma. In the case of the T-beam injection, in the range of parameters, for which the neutron yield is close to its maximum, the amount of tritium in the FC is lower than in the case of the D-beam injection. The neutron yield can be increased to 6.5 × 10¹⁷ n/s if full separation of the D and T is introduced for the gas pumped out from the divertor and puffed back into the torus. With this approach, in the case of the tritium beam, the amount of tritium in the FC is T_inv of ~170 g. If this approach is used in the case of the deuterium beam, the neutron yield can reach 7.0 × 10¹⁷ n/s. However, in this case, the amount of tritium contained in the FC increases to 215 g. The results of the analysis performed are used for optimizing the FC of the FNS-C (compact) fusion neutron source, which is planned for construction in the framework of the comprehensive program of the State Corporation Rosatom “Development of Engineering, Technology and Scientific Research in the Field of Using Atomic Energy in the Russian Federation for the Time Period up to 2030.”

The fuel cycle (FC) of a fusion reactor includes the following operations with hydrogen-containing gas mixtures: tokamak pumping, hydrogen isotope extraction from tokamak exhaust, tritium separation from hydrogen-containing impurities, separation of hydrogen isotopes, fuel injection into plasma, processing of tritium-containing radioactive waste. Processing and purification of fuel is a delicate and multistage process, the increased requirements for which are justified by considerations of radiation safety and economic efficiency. Fusion devices and hence FCs fusion differ significantly in scale, functional features, and amount and flux of tritium in systems, which makes it practically impossible to use the same technologies in different installations. This leads to the need to consider the possibility of using new technologies in FC systems and to find and develop systems based on efficient technologies for extracting hydrogen isotopes from gas mixtures. One such technology is the electrochemical hydrogen pump (EHP). There are three types of EHP based on solid oxide electrolyte (SOE), phosphate electrolyte (PHE) and solid polymer electrolyte (SPE). The article considers the possibility of using EHP in various FC systems for selective pumping of the fuel mixture, purification of the fuel mixture from impurities, and tritium separation from the breeder gas, as well as for other purposes.