Electrospun fiber-based photocatalysts demonstrate significant potential in addressing global environmental and energy challenges, primarily due to their high specific surface areas and unique properties. This review examines recent advances in the application of these materials in photocatalytic processes, with a particular focus on water splitting and hydrogen production. The principles of the electrospun method are described in detail, along with the operating parameters, material characteristics, and environmental conditions that affect the fiber formation. Additionally, the review discusses the challenges, advantages, and future prospects of photocatalysts incorporating carbon materials, metals, semiconductors, and hybrid structures with improved performance. These materials have the potential to significantly improve the efficiency of hydrogen energy production, water purification, and CO₂ recovery, highlighting their importance in engineering sciences.

This paper describes the results of investigations of 08Cr16Ni11Mo3 (AISI 316 steel analogue) austenitic stainless steel irradiated in BN-350 breeder reactor at irradiation conditions close to that for Light Water Reactor (LWR) Internals. The pores were found in 08Cr16Ni11Mo3 steel irradiated at temperature 280°C up to rather low damage 1.3 dpa and with dose rate 3.9×10−9 dpa/s. There were obtained dose rate dependencies of yield strength, ultimate strength and ductility for 08Cr16Ni11Mo3 steel irradiated up to 7–13 dpa at 302–311°C. These dependencies show a decrease in both yield strength and ultimate strength when dose rate decreases. There was observed an apparent decrease in total elongation when dose rate decreases, which was presumably connected with the pores formation in the steel at low dose rates.

This article presents the results of studies of the degradation of the structure and mechanical properties of the core materials BN-350 fast neutron and research WWR-K reactors required to justify the service life extension of early-generation power and research reactors. Extending the service life of nuclear reactors is a modern problem, since most operating reactors are early-generation reactors that have exhausted their design lifespan. The possibility of extending the service life is largely determined by the condition of the structural materials of the nuclear facility, i.e., their residual resources must ensure safe operation of the reactor. For the SAV-1 alloy, the structural material of the WWR-K reactor, studies were conducted on witness samples which were in the active zone during its operation for 56 years. It was found that yield strength and tensile strength of the irradiated SAV-1 alloy decreased by 24–48%, and relative elongation decreased by ~2% compared to the unirradiated alloy. Inside the grains and along their boundaries, there were particles of secondary phases enriched with silicon, which is typical for aged aluminum alloys. For irradiated structural steels of power reactors, studied at 350–450 C, hardening and a damping nature of creep were revealed, caused by dispersion hardening and the Hall–Petch effect.