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Recently, high entropy alloy (HEA) has become a research hotspot as a new candidate structural material in nuclear reactors due to its good irradiation resistance in swelling and hardening. Focusing on the temperature effect of irradiation damage, this work investigated the influence of irradiation temperature on dislocation evolution and irradiation hardening of HEAs. CoCrFeMnNi HEA was irradiated by high-energy Fe ions at room temperature and 500 °C. It was found that dense small dislocations were produced in the damage attenuation region (i.e., the tail of the ion range) of HEAs after irradiation at room temperature, whereas the irradiation-induced dislocations could not be observed in the damage attenuation region when the irradiation temperature was increased to 500 °C. For the small-sized dislocations, dissociation may occur more easily than long-range migration in HEAs (such as CoCrFeNi systems) due to the inhibition of defect migration and the decrease in defect binding energy, and this order is reversed in pure metals (such as Ni, W). Therefore, at 500 °C irradiation, small dislocations in the damage attenuation region of CoCrFeMnNi HEAs were dissociated before migrating to deeper regions, thereby resulting in the depth of dislocation distribution smaller than the stopping and range of ions in matter-calculated damage stopping depth, unlike the phenomenon in pure metals where dislocations migrated to regions exceeding the calculated depth. In addition, the dislocation density of CoCrFeMnNi HEAs decreased significantly due to the promotion of dissociation and merging of dislocations by elevated temperatures, and the hardening after 500 °C irradiation was less than that after room temperature irradiation. [ABSTRACT FROM AUTHOR]

Process intensification is crucial for industrial implementation of biocatalysis and can be achieved by continuous process operation in miniaturized reactors with efficiently immobilized biocatalysts, enabling their long-term use. Due to their extremely large surface-to-volume ratio, nanomaterials are promising supports for enzyme immobilization. In this work, different functionalized nanofibrous nonwoven membranes were embedded in a two-plate microreactor to enable immobilization of hexahistidine (His 6)-tagged amine transaminases (ATAs) in flow. A membrane coated with Cu2+ ions gave the best results regarding His 6 -tagged ATAs immobilization among the membranes tested yielding an immobilization yield of up to 95.3 % for the purified N -His 6 -ATA-wt enzyme. Moreover, an efficient one-step enzyme immobilization process from overproduced enzyme in Escherichia coli cell lysate was developed and yielded enzyme loads up to 1088 U mL−1. High enzyme loads resulted in up to 80 % yields of acetophenone produced from 40 mM (S)-α-methylbenzylamine in less than 4 min using a continuously operated microreactor. Up to 81 % of the initial activity was maintained in a 5-day continuous microreactor operation with immobilized His 6 -tagged ATA constructs. The highest turnover number within the indicated time was 7.23·106, which indicates that this immobilization approach using advanced material and reactor system is highly relevant for industrial implementation. [Display omitted] • A microreactor with nonwoven nanofiber membrane was developed. • Nanofibers with various functional groups were tested for enzyme immobilization. • His6-tagged amine transaminase was immobilized on membranes with metal ions. • One-step purification and immobilization with very high enzyme loads was reached. • Long-term stability of microreactor system and high turnover numbers were obtained. [ABSTRACT FROM AUTHOR]

Syngas, mostly hydrogen and carbon monoxide, has traditionally been produced from coal and natural gas, with biomass gasification later emerging as a renewable process. It is widely used in fuel synthesis through the Fischer–Tropsch (FT) process, where the H2/CO ratio is crucial in determining product efficiency and quality. In this sense, this study aimed to reform an emulated syngas resulting from the supercritical water gasification of biomass, tailoring it to meet the H2/CO ratio required for FT synthesis. Conditions resembling dry reforming were applied, using temperatures from 600 to 950 °C and steel wool as a catalyst. Additionally, the effects of Inconel and stainless steel as reactor materials on syngas reforming were investigated. When Inconel was used, H2/CO ratios ranged between 1.04 and 1.84 with steel wool and 1.28 and 1.67 without. When comparing reactions without steel wool performed either in the Inconel or the stainless steel reactors, those using Inconel consistently outperformed the stainless steel ones, achieving CH4 and CO2 conversions up to 95% and 76%, respectively, versus 0% and 39% with stainless steel. It was concluded that the Inconel reactor exhibited catalytic properties due to its high nickel content and specific oxides. [ABSTRACT FROM AUTHOR]

Nanoparticle‐stabilized, bicontinuous interfacially jammed emulsion gels (bijels) find potential applications as battery, separation membrane, and chemical reactor materials. Decreasing the liquid domain sizes of bijels to sub‐micrometer dimensions requires surfactants, complicating bijel synthesis and postprocessing into functional nanomaterials. This work introduces surfactant‐free bijels with sub‐micrometer domains, solely stabilized by nanoparticles. To this end, the covalent surface functionalization of silica nanoparticles is characterized by thermogravimetric analysis, mass spectrometry, Fourier‐transform infrared spectroscopy, and contact angle measurements. Bijels are generated with the functionalized nanoparticles via solvent transfer induced phase separation (STrIPS), enabling the optimization of nanoparticle functionalization and surface ionization. Nanoparticles of intermediate functionalization and controlled negative surface charge stabilize bijels with sub‐micrometer liquid domains. This remarkable control over bijel synthesis provides urgently needed progress to facilitate the widespread implementation of bijels as nanomaterials in research and applications. [ABSTRACT FROM AUTHOR]

Reduced activation ferritic steels are being explored as possible cladding tube materials for nuclear reactors because of their low activation and excellent irradiation resistance. In the current investigation, reduced activation ferritic steel (Fe–15Cr–2W) is processed by mechanical alloying of elemental powders followed by hot powder forging. Mechanical alloying is carried out in a Simoloyer attritor mill (Zoz GmbH), after which the powders are placed in a mild steel can and forged at 1200 °C in H2 atmosphere. X‐ray diffraction and transmission electron microscopy (TEM) investigation reveal that 10 h of mechanical alloying is required to achieve complete dissolution of Cr and W in the Fe matrix powder. The relative density and hardness distribution of the forged slab is evaluated in longitudinal as well as transverse direction to optimize the powder forging operation. Electron backscatter diffraction analysis showed dynamic recrystallization to take place during the course of hot powder forging. Tensile tests are performed at room temperature as well as at elevated temperatures (600 and 700 °C). The yield strength and ultimate tensile strength at room temperature as well as at elevated temperatures are found to be higher than those reported in literature for reduced activation ferritic steels consolidated by other techniques. [ABSTRACT FROM AUTHOR]

Zirconium alloys are widely used in nuclear water reactors as cladding materials. The cladding materials will absorb hydrogen from high temperature water during the operation of nuclear reactor. In cladding tubes, it has been common sense that circumferential hydrides form without stress, while radial hydrides can form when the hydrides are reoriented under stress loading. In this study, we found that a high heating rate can result in hydride reorientation behavior even without stress. At elevated heating rates, the zirconium alloy clad tube developed a non-uniform strain gradient along the direction of heat conduction. Hydrogen atoms migrate preferentially to areas of elevated stress and precipitate as hydrides that are perpendicular to the direction of tensile stress, resulting in the formation of radial hydrides that appear as "sun spots" macroscopically. Additionally, the high heating rate disrupts the {0001}α∥{111}δ, <11–20>α∥<110>δ orientation relationship between the hydride and the substrate, which potentially facilitates crack propagation. [ABSTRACT FROM AUTHOR]

Recent experimental investigations have demonstrated that a substantial amount of H−/D− ions can be formed by thermal desorption processes. Based on these new findings, new mini axial and coaxial-type neutron tubes have been developed for the production of high or low-energy neutrons via the d-d, d-10B, d-7Li or p-7Li nuclear reactions. By operating these mini neutron tubes with a high frequency AC high-voltage supply, short pulses of high intensity neutron beams can be generated. Multiple applications, such as carbon and well logging, neutron imaging, cancer therapy, medical isotope production, fission reactor start-up, fusion reactor material evaluation, homeland security and space exploration can be performed with the subcompact neutron generator system. It is shown that the performance of these new mini neutron tubes can exceed those of the conventional plasma-based neutron sources. [ABSTRACT FROM AUTHOR]

The Ti-Nb-Fe-Cr alloys are used in various fields, such as nuclear radiation shielding, cladding material in nuclear reactors, and implants in the medical field. It is one of the best materials for biomedical applications as it is biocompatible, is corrosion resistant, and has good mechanical properties. As nuclear radiation emanates from various sources in a nuclear reactor, the behavior of this alloy with the interaction of gamma and neutrons has not been well studied. In the present investigations, the interaction of gammas and neutrons with Ti-27Nb-7Fe-xCr (x = 0, 2, 4, 6, and 8 wt%) alloys is studied in order to understand the radiation shielding properties and their usefulness in biomedical applications. Gamma-ray–interaction parameters such as MAC, HVL, MFP, Zeff, Zeq, Neff, and multilayer energy absorption buildup factor (MLEABF) are estimated using EpiXs, PyMLBUF, and NGCal software in the energy range of 1 keV to 15 MeV. The multilayer buildup factor (MLBF) is calculated for cortical bone and for alloys with varying Cr concentrations. Comparison of the MLBF values of alloys with cortical bone shows that in the lower-energy region as well as the higher-energy region above 0.5 MeV, alloys and cortical bone yield the same values, indicating that the alloys behave as cortical bone in this energy region. Mass attenuation factors (MAFs) of thermal and fast neutrons are also calculated for various elastic modulus values of selected alloys at thermal and fast neutrons. It is found that the elastic modulus increases with increasing MAF values of both fast and thermal neutrons. By increasing the Cr content in the Ti-27Nb-7Fe alloy, the elastic modulus decreases. The relationship between the MAF of neutrons and the elastic modulus of the alloy is established for the first time. [ABSTRACT FROM AUTHOR]

Under the goals to reach carbon peaking and carbon neutrality, the clean and efficient utilization of fossil fuels has become a crucial aspect of the energy transition. Methane as one of the cleanest fossil energies, has been widely regarded as a key technology with good development prospects in energy transition for high-value and clean utilization. Methane pyrolysis for hydrogen process has the advantages of producing high-purity H2 and carbon materials without directly generating CO2. Three typical methane pyrolysis for hydrogen processes: catalytic methane pyrolysis in fluidized beds, plasma methane pyrolysis, and molten medium methane pyrolysis were summarized and their principles, advantages, disadvantages and research progress were introduced. Among these processes, catalytic methane pyrolysis in fluidized beds requires the least energy but the catalysts are prone to carbon deposition, leading to unsustainable reactions and difficulty in separating carbon materials from catalysts, affecting downstream use. Plasma methane pyrolysis has a high conversion rate but requires a substantial energy input. Molten medium methane pyrolysis has lower energy consumption, and the carbon materials float on the surface of the molten medium, not hindering the continuous reaction and are easily separated and collected. However, the high temperature of the molten medium requires high requirements for the reactor materials, and the cost of purifying carbon materials also affects the economic viability of this process for industrialization. Combining renewable energy is the future trend towards achieving clean and efficient methane pyrolysis for hydrogen. Utilizing renewable energy sources such as solar and wind energy to provide the energy required for methane pyrolysis processes, can reduce the reliance on traditional energy. From current research and development trends, molten medium methane pyrolysis process for hydrogen integrated with renewable energy sources is expected to become the key process for methane pyrolysis for hydrogen and carbon materials. [ABSTRACT FROM AUTHOR]

In order to promote the application of SiC ceramics in nuclear reactor cladding materials, the preparation and application of (Zr-20Cu)-10Nb(wt.%) filler metal for SiC brazing were studied. The microstructure morphologies, phase composition and wettability to SiC ceramics of the filler metal prepared by the suction casting method were evaluated. The results show that the filler metal mainly consists of Cu, β-Zr and a diffusely distributed second phase β-Nb, and it can well wet SiC ceramics. In addition, the effects of brazing temperature (1000-1200 °C) and holding time (10-20 min) on the phase composition, microstructure and properties of SiC joints brazed by Zr-Cu-Nb filler metal were investigated, respectively. During the brazing process, the interfacial reactions occur between the filler metal and SiC to produce ZrC, Zr2Si and CuZr, and increasing the brazing temperature or holding time will promote the interfacial reactions. The effects of brazing temperature on joint properties are more significant compared to holding time. The shear strength of brazed SiC ceramic joints first increases and then decreases with the increase of brazing temperature or holding time, and the maximum shear strength of 101 MPa could be achieved when brazing at 1100 °C for 15 min. [ABSTRACT FROM AUTHOR]

Damages induced on the tungsten surface at two different operating conditions of a kilojoule plasma focus device are studied. In one condition, the tungsten samples were exposed to axial plasma shocks that are formed after pinch disruption, and in the other condition, the pinch phenomenon was absent or weak. Melting, craters, and cracking on the surfaces were observed in both cases. In the former case, the charged particle beams and post-pinch material ejection will play a role in impacting the surface; however, in the latter case those phenomena will have small contributions because of the absence or weak formation of the pinch. A damage factor of ∼ 10 9 W m−2 s0.5 was estimated at a distance of 3 cm from the pinch exit using the method given in Akel et al. [J. Fusion Energy 35, 694–701 (2016)] and Klimov et al. [J. Nucl. Mater. 390, 721–726 (2009)] for the former case. The present work suggests that at pressures lower than the pinch-occurring pressure, only axial plasma shock effects on the targeted surface can be studied and that they can be separated from the effects produced by the charged particle beams mixed with axial plasma shocks in the case of pinch occurrence. [ABSTRACT FROM AUTHOR]

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To expedite enrichment of anaerobic ammonia-oxidizing bacteria (AnAOB) as a way to reduce the start-up time, leading to a quicker transition into stable operation, the anaerobic ammonia oxidation (anammox) process was initiated by a biofilm reactor with polyurethane porous material. The enrichment of anammox bacteria was studied by progressively increasing the influent substrate concentration while simultaneously decreasing hydraulic retention time. Following a 73 d start-up and subsequent 103 d enrichment phase, the removal rates of ammonia and nitrite reached 97.87% and 99.96%, respectively, and the community was characterized by the development of brick-red anammox biofilms and granules. The predominant bacterial phyla within the reactor were Planctomycetota, Chloroflexi, and Proteobacteria, with relative abundances of 25.25%, 29.41%, and 14.3%, respectively, and the dominant genus was Candidatus brocadia, comprising 20.44% of the microbial community. These findings indicate that the polyurethane porous material biofilm reactor is conducive to the enrichment of AnAOB. After enrichment, the anaerobic microbial community exhibited significant richness and diversity, with anammox bacteria as the primary group. [ABSTRACT FROM AUTHOR]

The ability to distinguish multiple forms of plutonium from one another, such as oxide and metal, is paramount in areas of nuclear nonproliferation and international safeguards. In its metal form, plutonium can be readily used in a nuclear weapon, while oxide forms are associated with nuclear reactor fuel. Oxide-based plutonium forms emit neutrons with an energy spectrum that is significantly different from the fission neutrons that are emitted from plutonium metal. Organic scintillation detectors output pulses that are proportional to the neutron energy deposited, and therefore present a means of distinguishing these plutonium forms based on their energy spectra. In this work, metal and oxide forms of plutonium were measured using a handheld detection system based on an organic glass scintillator. Monte Carlo modeling of these experiments was performed to provide insight into the origin of the features in the observed light output spectra. Through analysis of multiple regions of these spectra, in a matter of minutes we were able to unambiguously discriminate oxide and metal plutonium forms from one another and from a plutonium-beryllium neutron source, which was considered for comparison because these sources are commonly used in industrial applications. The ability to discriminate weapons-usable material from nuclear reactor fuel has applications in nuclear treaty verification and safeguards. [ABSTRACT FROM AUTHOR]

Multi-principal element alloys demonstrate high strength, thermal stability, and irradiation resistance, making them excellent candidate materials for applications in nuclear reactors and other harsh environments. Some studies have examined the use of high-pressure torsion to strengthen MPEAs through grain size reduction and strain hardening. However, no studies have investigated the effect of HPT on secondary phases (precipitates) within an MPEA. Two alloys, (Fe0.3Ni0.3Mn0.3Cr0.1)94Ti2Al4 containing Ni(Ti, Al) B2 phase, and CrFe σ phase, and single-phase Fe0.3Ni0.3Mn0.3Cr0.1, were fabricated by casting and heat treatment. Both alloys were then processed with HPT to study microstructural evolution. Scanning electron microscopy (SEM) and transmission electron microscopy (TEM) were used to characterize the alloys before and after HPT processing. HPT processing produced a nanocrystalline structure in both alloys, but (Fe0.3Ni0.3Mn0.3Cr0.1)94Ti2Al4 exhibited a significantly smaller grain size and higher dislocation density than Fe0.3Ni0.3Mn0.3Cr0.1, with corresponding higher hardness. Before HPT, the (Fe0.3Ni0.3Mn0.3Cr0.1)94Ti2Al4 alloy consisted of large grain (~ 400 μm) and precipitates, including B2 of ~ 38 μm average size, B2 of ~ 0.7 μm average size, and small amounts of σ of ~ 1.5 µm average size. After HPT, the larger B2 precipitates were decreased in size and volume fraction, while the smaller B2 precipitates were completely dissolved; the σ precipitates appeared unaffected by HPT, likely due to their much higher hardness. Observation of the B2 precipitate distribution along radial distance indicates that the strain caused the precipitates to fracture at intermediate strain (γ = 125) and dissolve at high strain (γ = 280). [ABSTRACT FROM AUTHOR]

Nuclear accidents, such as the Fukushima nuclear accident, have highlighted the necessity for accident‐tolerant fuel (ATF) cladding. Previous studies focused on coating the outside of Zr alloy currently used in nuclear reactors with an oxidation‐resistant material in a vacuum environment. This limits the coating to the inside of the cladding and does not tend to achieve a uniform coating on meter scale cladding. In this study, a room temperature and non‐vacuum‐based swaging–drawing process was demonstrated as an alternative cladding manufacturing process. It enables both the inner and outer sides of the 2‐m‐long Zr alloy cladding to be uniformly covered with a 100‐μm‐thick corrosion‐resistant material (316‐L stainless steel; SS316L), thereby minimizing its high‐temperature oxidation and avoiding failures. After the swaging–drawing process, there was a gap of less than 1 µm between outer SS316L and Zr alloy and a gap of about 12 µm between inner Zr alloy and SS316L. The high‐temperature oxidation properties of the resulting triplex Gachon ATF cladding tube (G‐tube) were evaluated up to 1,200°C in an atmospheric environment. Following heat treatment at 1,200°C, the control cladding completely oxidized and ruptured, potentially causing leakage of radioactive material during application. In contrast, only 15% of the G‐tube cladding manufactured by the swaging–drawing process was oxidized despite a gap, and the Zr alloy of the G‐tube changed phase from α‐Zr to α‐Zr (O) and prior β‐Zr. The cladding microstructure, oxide layer, and oxidation mechanism were analyzed through microscopy, X‐ray diffraction, and thermogravimetric analysis. As a result, it was confirmed that SS316L completely prevented oxygen diffusion into the bulk Zr alloy. In addition, there was no elemental diffusion between SS316L and the Zr alloy. These results demonstrate the feasibility of using room temperature, nonvacuum environment‐based swaging–drawing process to fabricate structurally stable ATF cladding at extremely high temperatures. [ABSTRACT FROM AUTHOR]

The works in the field of radiation materials science of target materials of neutron sources based on the subcritical assemblies controlled by linear accelerators of electrons or protons, so-called accelerator driven systems (ADS), are reviewed. Now, electronuclear ADS systems are the prototype of safe nuclear reactors of the 5th generation. In connection with the physical start-up of the neutron source facility of the National Scientific Centre 'Kharkiv Institute of Physics and Technology' of the N.A.S. of Ukraine (NSC 'KhIPT' NASU), the target of which is fabricated from powdered tungsten covered with tantalum, the issues of preparation technology, construction, and physical and mechanical properties of W-Ta materials of targets in non-irradiated and irradiated states are considered. The nuclear-physical processes of radiation damage to the target during co-irradiation of it with both neutrons of a subcritical assembly and high-energy e- and g-beams are analyzed. The target resources of ADS systems are estimated. As noted, the difficulty of predicting the 'survivability' of the W-Ta target also relates to the fact that, except for the works carried out at the NSC 'KhIPT' NASU in 70-90th of the last century, there are no experimental works in the world on the radiation damage of reactor materials by high-energy electrons with an energy of 100 MeV and above. [ABSTRACT FROM AUTHOR]

Blister formation has been an emerging research topic for extreme ultraviolet (EUV) mirrors exposed to hydrogen plasmas. Similar to plasma-facing materials in nuclear fusion reactors, it has been reported that blister formation in EUV mirrors is initiated by hydrogen uptake due to hydrogen ion or atom bombardment. However, the research so far has focused on Mo/Si multilayers exposed to only hydrogen ions or atoms, while the EUV mirror typically has a Ru capping layer facing hydrogen plasmas. We present experimental work to measure plasma-induced hydrogen uptake of Ru films. We bombarded our designed Ru-capped target with a low-temperature deuterium plasma and measured the deuterium retention using elastic recoil detection. Contrary to ion-driven deuterium uptake, the deuterium uptake rate of the Ru film had no dependence on the deuterium ion flux or energy after a period of plasma exposure. A reaction–diffusion model has been built to calculate the time evolution of deuterium retention, which well fits the experimental data. Based on this model, we conclude that the surface composition of the Ru film is the limiting factor for the deuterium uptake, which is seriously weakened when the surface is covered by Ru oxide. After the Ru oxide is reduced by the plasma, the uptake rate is predominantly driven by the deuterium surface coverage on metallic Ru. Our model also indicates that at the deuterium-populated Ru surface, deuterium has a low absorption barrier to penetrate the surface, which is supported by previously reported computational work. [ABSTRACT FROM AUTHOR]

The stability of structural materials in extreme nuclear reactor environments—with high temperature, high radiation, and corrosive media—directly affects the lifespan of the reactor. In such extreme environments, an oxide layer on the metal surface acts as a passive layer protecting the metal underneath from corrosion. To predict the irradiation effect on the metal layer in these metal/oxide bilayers, nondestructive depth-resolved positron annihilation lifetime spectroscopy (PALS) and complementary transmission electron microscopy (TEM) were used to investigate small-scale defects created by ion irradiation in an epitaxially grown (100) Fe film capped with a 50 nm Fe2O3 oxide layer. In this study, the evolution of induced vacancies was monitored, from individual vacancy formation at low doses—10−5 dpa—to larger vacancy cluster formation at increasing doses, showing the sensitivity of positron annihilation spectroscopy technique. Furthermore, PALS measurements reveal how the presence of a metal–oxide interface modifies the distribution of point defects induced by irradiation. TEM measurements show that irradiation induced dislocations at the interface is the mechanism behind the redistribution of point defects causing their accumulation close to the interface. This work demonstrates that the passive oxide layers formed during corrosion impact the distribution and accumulation of radiation induced defects in the metal underneath and emphasizes that the synergistic impact of radiation and corrosion will differ from their individual impacts. [ABSTRACT FROM AUTHOR]

In this study, a new and facile process was developed for the preparation of composite catalysts based on tungsten oxide (WO3) by batch reactor routes. The structures, morphologies, compositions, and characteristics of synthesized materials were investigated and confirmed. Using batch reactor processes, WO3 nanorods (WO3 NR), heterostructures of WS2/WO3 nanobricks (WS2/WO3 NB), and WS2/WO3 nanorods (WS2/WO3 NR) were successfully prepared. The prepared materials were then employed for hydrogen evolution reaction (HER) to investigate their catalytic performance. The results indicated that the electrocatalytic activities of WS2/WO3 NR are significantly improved compared to those of WO3 NR and WS2/WO3 NB. This improvement could be attributed to the formation of heterostructure between WS2 and WO3 elements in highly uniform materials, which could create the synergistic effect and further improve the catalytic activities of the catalyst. The data shows that the Tafel slope of WS2/WO3 NR (82.7 mV dec−1) is significantly lower than that of WO3 NR (112.5 mV dec−1) and WS2/WO3 NB (195.5 mV dec−1). Furthermore, the resistance of WS2/WO3 NR (397.7 Ω) is markedly decreased compared to those of WO3 NR (1816 Ω) and WS2/WO3 NB (3597 Ω). The results indicate that WS2/WO3 NR could be a great catalyst for electrochemical applications. [ABSTRACT FROM AUTHOR]

The use of lipase immobilized on an octyl-agarose support to obtain the optically pure enantiomers of chiral drugs in reactions carried out in organic solvents is a great challenge for chemical and pharmaceutical sciences. Therefore, it is extremely important to develop optimal procedures to achieve a high enantioselectivity of the biocatalysts in the organic medium. Our paper describes a new approach to biocatalysis performed in an organic solvent with the use of CALB-octyl-agarose support including the application of a polypropylene reactor, an appropriate buffer for immobilization (Tris base—pH 9, 100 mM), a drying step, and then the storage of immobilized lipases in a climatic chamber or a refrigerator. An immobilized lipase B from Candida antarctica (CALB) was used in the kinetic resolution of (R,S)-flurbiprofen by enantioselective esterification with methanol, reaching a high enantiomeric excess (eep = 89.6 ± 2.0%). As part of the immobilization optimization, the influence of different buffers was investigated. The effect of the reactor material and the reaction medium on the lipase activity was also studied. Moreover, the stability of the immobilized lipases: lipase from Candida rugosa (CRL) and CALB during storage in various temperature and humidity conditions (climatic chamber and refrigerator) was tested. The application of the immobilized CALB in a polypropylene reactor allowed for receiving over 9-fold higher conversion values compared to the results achieved when conducting the reaction in a glass reactor, as well as approximately 30-fold higher conversion values in comparison with free lipase. The good stability of the CALB-octyl-agarose support was demonstrated. After 7 days of storage in a climatic chamber or refrigerator (with protection from humidity) approximately 60% higher conversion values were obtained compared to the results observed for the immobilized form that had not been stored. The new approach involving the application of the CALB-octyl-agarose support for reactions performed in organic solvents indicates a significant role of the polymer reactor material being used in achieving high catalytic activity. [ABSTRACT FROM AUTHOR]

15 Cr ferritic oxide dispersion strengthened (ODS) steels are considered prime fuel cladding materials in nuclear reactors due to their excellent creep, swelling, and oxidation resistance. In the present study, the nominal compositions Fe–15Cr–2W–xY2O3 (x = 0, 0.3, 0.7, and 1.0) of ferritic ODS steels were prepared by mechanical alloying followed by spark plasma sintering. The sintered samples were annealed at different temperatures of 950, 1100, and 1250°C with a holding time of 60 min at respective temperatures. Further, the samples were also annealed at 1100°C for various durations of 0, 60, and 120 min. The role of varying yttria dispersoids and annealing temperatures on the grain growth kinetics, as well as their mechanical properties (hardness and compressive strength), were analyzed. The compressive strength of the sintered samples with varying yttria contents and at elevated temperatures of 600 and 700°C was determined. Modeling of compressive yield strength at room and elevated temperatures, as well as a correlation with the experimental values, were established for all the compositions. The grain growth exponent (n) and activation energy (Q) rose with the increase in yttria content and were estimated to be 11.52 and 612.91 kJ/mol, respectively, with 1.0 wt.% yttria. The grain size was nearly stable at the annealing temperature of 1100°C. A significant rise in compressive strength at room temperature and elevated temperatures was observed with a yttria reinforcement content of 0.7 wt.%. According to the strength model at different conditions, the role of ultrafine grains and dispersoids seemed to be predominant at room temperature and high temperatures, respectively. [ABSTRACT FROM AUTHOR]

This study evaluated the gamma and neutron-shielding parameters, electron and charge particles' interaction properties of three different stainless steel-based alloys (2507SS, 304SS and AISI1018) with densities of 7.82, 7.81 and 7.82, respectively for deployment in nuclear reactor construction. Gamma shielding parameters were evaluated with Phy-X/PSD, Py-MLBUF and GRASP computer programs within the energy range of 0.2 MeV to 15 MeV. The relative deviations (RD %) between two of these computational platforms were calculated. Fast Neutrons Effective Removal Cross-Sections (FNRCS) and Macroscopic Removal cross-section (MRCS) values for fast neutrons were calculated with Phy-X/PSD and MRCS software, while thermal/fast neutron attenuation factors were calculated with NGCals software. The continuous-slowing-down approximation (CSDA) range and total stopping power (TSP) values of H+ and He++ ions were estimated with SRIM Monte Carlo code in a wide energy range of 1–20 MeV. The projectile range of electrons was computed with ESTAR NIST software within the kinetic energy of 0.01–1000 MeV. Gamma-ray, thermal neutron and fast neutron transmission factors (TFs) values were calculated for all samples for a range of well-known energies. From the obtained results, all evaluated parameters were dependent on the composition of shielding materials, the type of radiation and the photon energy of the radiation. The maximum MAC values obtained from the three alloy samples are 0.1505 cm2/g for 2507SS, 0.1453 cm2/g for 304SS and 0.1459 cm2/g for AISI1018 at 0.02 MeV. These results projected 2507SS as a better shielding alloy when compared with other investigated alloy samples. Considering the closeness of densities of investigated alloy samples the shielding and interaction properties of all alloy samples gave excellent and similar results implying that all investigated alloys will effectively serve as a radiation-shielding material in a nuclear reactor. [ABSTRACT FROM AUTHOR]

A differential scanning calorimeter was designed to study the thermal analysis of the materials for the nuclear reactor/allied facilities. The hardware and software were developed for measuring the calorimetric signals and had various features for the thermal analysis of the material. The processing of calorimetric signals was carried out by a programmable system-on-chip. Using a programmed temperature profile, the differential scanning calorimeter setup was tested, calibrated, and validated with standard aluminum and indium metal samples. It was found that the obtained results agree with the literature values. The methodology for the experiment was optimized. Various experiments were carried out, and the thermal analysis of different materials for nuclear reactors was investigated and studied. The instrumentation is also compact, accurate, reliable, and cost-effective for the thermal analysis of materials. [ABSTRACT FROM AUTHOR]

This study reports on the magnetic hardening phenomenon of heavily helium ion-irradiated iron–chromium alloys. The alloys are important structural materials in next-generation nuclear reactors. In some cases, problems may arise when the magnetic properties of the materials change due to neutron irradiation. Therefore, it is necessary to understand the effects of irradiation on magnetism. Helium irradiation was conducted as a simulated irradiation, and the effect of cavity formation on magnetic properties was thoroughly investigated. High-quality single-crystal Fe-x%Cr (x = 0, 10, 20) films, with a thickness of 180–200 nm, were fabricated through ultra-high vacuum evaporation. Subsequently, irradiation of 19 dpa with 30 keV He+ ions was conducted at room temperature. X-ray diffraction measurements and electron microscopy observations confirmed significant lattice expansion and the formation of high-density cavities after irradiation. The magnetization curve of pure iron remained unchanged, while magnetic hardening was noticed in iron–chromium alloys. This phenomenon is believed to be due to the combined effect of cavity formation and changes in the atomic arrangement of chromium. [ABSTRACT FROM AUTHOR]

In this paper, a series of novel trans‐1, 4‐polyisoprene/isotactic polybutene (TPI/iPB) in‐reactor blends were synthesized by isoprene and butene sequential two‐stage polymerization technology with spherical TiCl4/MgCl2 type Ziegler‐Natta catalyst. The components, structures, and properties of the as‐obtained TPI/iPB reactor blends were characterized by gel permeation chromatography, Fourier transform Infrared spectroscopy, and differential scanning calorimetry. The active trans‐1, 4‐polyisoprene (TPI) particles obtained in the initial isoprene polymerization by the Z‐N catalyst can be acted as microreactors to initiate butene polymerization subsequently. The TPI/iPB reactor blends with varied components were in situ synthesized within the reactor. The preparative‐temperature rising elution fractionation (p‐TREF) technique was used to fractionate the TPI/iPB reactor blends based on the elution temperature ranged from −40°C to 90°C. The weight distribution and microstructure of each fraction were investigated. The reactor blends are composed of crystallizable high trans‐1, 4‐uint polyisoprene obtained from the first‐stage isoprene polymerization, high isotactic polybutene obtained from the second‐stage butene polymerization and TPI‐b‐iPB block copolymer with different sequence structure obtained from the initial time of the second stage. This work is expected to propose the possible polymerizations of a‐olefins and conjugated dienes by using heterogeneous Ziegler‐Natta catalyst and provide a kind of novel rubber/plastic reactor blend materials. Highlights: The first synthesis of the trans‐1, 4‐polyisoprene/isotactic polybutene reactor blends.p‐TREF technique was used to analyze the composition of TPI/iPB reactor blends.Detailed structure characterization of the fractions including GPC and NMRTPI‐b‐iPB block copolymer can be obtained through the sequential polymerization.Rubber/plastic reactor blend materials were fabricated by the Z‐N catalyst. [ABSTRACT FROM AUTHOR]

The present research work describes the effect of normalizing and tempering (N&T) treatment on microstructure evolution in various zones of gas tungsten arc-welded (GTAW) P92 pipe weldments. For N&T treatment, P92 pipe weldments are subjected to various normalizing (950−1150°C) and tempering (730−800°C) temperatures. The effect of varying heat treatment on tensile properties and hardness of P92 pipe weldments are studied for V-groove and narrow-groove weld designs. The effect of increase in normalizing temperature (at fixed tempering temperature) results in increase in strength and hardness, while increase in tempering temperature (at fixed normalizing temperature) results in the decrease in strength and hardness of P92 steel weldments. Creep strength-enhanced ferritic/martensitic P92 steel is considered as a candidate material for the reactor pressure vessels and reactor internals of Very High Temperature Reactor (VHTR). The heterogeneous microstructure formation across the P92 weldments leads to premature Type IV cracking and makes the weldability of P92 steel as a serious issue. The better combination of strength, ductility and microstructure are obtained for the maximum normalizing temperature of 1050°C and tempering temperature of 760°C. The effect of increase in normalizing temperature (at fixed tempering temperature) results in increase in strength and hardness, while increase in tempering temperature (at fixed normalizing temperature) results in the decrease in strength and hardness of P92 steel weldments. [ABSTRACT FROM AUTHOR]

The available literature experimental data on corrosion resistance of traditional and advanced fuel rod cladding materials for water-cooled reactors are summarized. A review of zirconium alloys, which have proven themselves in operation for more than half a century, is presented. As noted, the research work is constantly being carried out to improve zirconium alloys by optimizing their composition, in particular, the amount of tin, niobium, iron and oxygen, as well as development of the new alloys. First of all, the direction of these works is stimulated by stringent nuclear energy requirements, including maximum safety, efficiency and environmental friendliness. At the same time, in the last decade, one of the main goals of researchers around the world is the development of nuclear fuel systems, which tolerate severe accidents. Another trigger for this was the accident in 2011 in Japan at the Fukushima-1 NPP. As the most optimal possible solution, it is considered the surface modification of zirconium alloys by the development of chromium coatings. Such coatings provide an increased corrosion resistance and wear resistance, as well as hydrogen pickup reduced at operating temperatures of the primary coolant and in emergencies. A more radical way to increase the fuel rod cladding accident resistance is to replace the zirconium alloy with another one. The best candidates are FeCrAl alloys and duplex stainless steels (DSS), whose corrosion resistance can be 50 times greater than that of zirconium alloys in loss-of-coolant accident (LOCA) conditions. Unfortunately, under the nominal water-cooled reactor operating conditions, a long-term operation such as claddings will lead to the corrosion product formation and its removal to the coolant followed by their activation and formation of deposits in the core and steam generator. This will certainly entail an increase in the radiation-dose rate from the primary circuit equipment. Considering the traditional and advanced water-cooled reactors claddings, which tolerate severe accident scenarios, an optimized zirconium alloy with chromium coating can be considered as the most advanced one. The corrosion resistance of such claddings is at least five times higher compared to traditional zirconium alloys both under normal operating conditions and severe accidents, and will not cause significant neutron absorption, coolant activation or deposit formation in the primary circuit. [ABSTRACT FROM AUTHOR]

Transitioning from fossil fuels to electricity as an energy source for industrial production is a crucial step toward future decarbonization. This study proposes a novel integrated electric heating hydrogen production foam catalyst and systematically investigates the influence of slurry composition and sintering temperature on the carrier strength. The research successfully obtains a catalyst with the required strength and significantly enhances the catalyst loading capacity. The findings indicate that: 1) Increasing the calcination temperature above 1200 °C can enhance foam strength and reduce cracks and voids in foam structure and joints. 2) Ceramics are more advantageous for the coating of the catalyst, enabling a catalyst loading of up to 152.11 g/L, which is 2.98 times higher than the maximum reported loading for electrically heated catalysts, the shedding rate is only 1.86 %. 3) The integrated foam catalyst is heated by an internal heating wire through in-situ Joule heating, reaching 650 °C in just 83 s. This approach, when compared to an external heating furnace, results in a 96.8 % reduction in start-up time, a 91 % reduction in total energy consumption. The integrated electrically heated foam enables a straightforward high-temperature insulation seal, making it suitable for various reactor materials and providing valuable insights for the broader application of electric heating. • A novel integrated electric heating hydrogen production foam catalyst is proposed. • The loading capacity of this catalyst is 2.98 times higher than previously reported for electrothermal catalysts. • The integrated electric heating catalyst achieves a temperature of 650 °C within 83 s. • In comparison with the external heating electric furnace, the total energy consumption is reduced by 91 %. [ABSTRACT FROM AUTHOR]

A high-temperature (up to 1600°C) installation for measuring the coefficient of linear expansion (CLE) using a relative method is described. The measuring unit is installed in a protective glove box, which allowed measurements to be performed on samples irradiated in a reactor. Changes in the length of the sample during heating were recorded with a clock-type indicator head with an accuracy of 1 μm with the measurement range of 0–10 mm. The installation was used to determine the values of swelling of irradiated samples during high-temperature annealing and to obtain the values of the CLE of promising reactor materials. The average relative measurement error is 8–11%. [ABSTRACT FROM AUTHOR]

Fundamental studies on screening catalyst materials with potential for application in large‐scale industrial ammonia reforming processes pose a significant challenge. Not only are the requirements on the catalyst itself critical, but the industrially relevant reaction conditions impose very high demands on the reactor materials, especially in terms of temperature and corrosion resistance, as well as the general layout of a suitable test rig on laboratory scale. Here, an automatized setup for high‐throughput studies on heterogeneous catalysts in ammonia decomposition, with a strong focus on reactor material selection and the capability of the setup to generate highly reproducible and thus comparable results, is described. With this approach, it is ultimately possible to contribute to identify possible catalyst materials applicable in industrial ammonia reforming. [ABSTRACT FROM AUTHOR]

The dose rate effect significantly impacts accelerated assessment experiments of the radiation resistance of critical structural materials in nuclear reactors. This study investigated the role of the dose rate on the irradiation hardening and defect cluster formation in a low-Cu reactor pressure vessel (RPV) steel A508-3 under 3.5 MeV Fe13+ ions irradiation at 290 °C, with chosen dose rates of 2.8 × 10–5, 1.4 × 10–4 and 2.8 × 10–4 dpa/s. Nano-indentation and atom probe tomography (APT) techniques were employed to characterize the mechanical property changes and defect clustering behavior, respectively. Nanoindentation results revealed that for a fixed dose, greater hardening was observed in the sample irradiated at a lower dose rate, and conversely, less hardening at a higher dose rate. APT results for samples with 3.0 dpa demonstrated that defect clusters exhibited characteristics of small size and low density under irradiation at a dose rate of 2.8 × 10–4 dpa/s, and large size and high density at a dose rate of 2.8 × 10–5 dpa/s. The ratio of solutes to Fe elements in a cluster also varied with changes in dose rate. The change in hardening, cluster size, number density and cluster composition suggested that the evolution of defect clusters was primarily driven by ballistic mixing, rather than radiation-enhanced diffusion, under heavy ions irradiation within the dose rate range of 2.8 × 10–5–2.8 × 10–4 dpa/s. [ABSTRACT FROM AUTHOR]

As one of the most important routes for the preparation of high-temperature materials, self-propagating metallurgy has clear advantages in preparing high melting point alloys containing titanium, tungsten and other components. In this study, the effects of basic reaction thermodynamics, composition and proportion of reducing agent, slagging agent addition, and other factors regarding the rate and stability of self-propagating reactions in the process of preparing TiAl alloy by self-propagating metallurgy were examined. Also, the stable self-propagating heat transfer process was calculated. The results showed that stable reaction conditions for preparing TiAl alloy by self-propagating metallurgy were 3200 J/g of heat effect per unit mass using an Al-Ca compound reducing agent, in which Ca accounts for 15 wt.% of the reducing agent. After the reaction was completed, a wrapped structure of alloy surrounded with slag was obtained. The slag composition was mainly (CaO)12 (Al2O3)7 and the alloy contents of Ti, Al, and O at 52.00, 47.90 and 0.269 wt.%, respectively. Moreover, stability was mainly related to the physicochemical properties of the feedstock, reactor specifications and feeding rate. The approximate feeding rate range can be calculated for different specifications of reactors and raw material composition, which provided data support for potential industrialization. [ABSTRACT FROM AUTHOR]

The article focuses on the environmental damage caused by the destruction of the Kakhovka Dam in southern Ukraine during the ongoing war, with the release of 18 cubic kilometers of water impacting ecosystems, wildlife, and human settlements along the Dnipro River. It mentions researchers highlight the challenges in assessing the environmental toll amid the conflict, including damage to a sturgeon breeding facility.

As a new type of nuclear reactor, the Gen-IV reactor attracts increasing interest from the industry. Tritium is produced in lead-cooled fast reactors (LFR), sodium-cooled fast reactor (SFR), and other Gen-IV reactors. It is necessary to study the tritium permeation behavior of 15-15Ti stainless steel as a candidate cladding and structural material in Gen-IV reactors. In this study, gas-driven permeation experiments were conducted to measure the high-temperature deuterium transport behavior parameters of 15-15Ti stainless steel. The results showed that the deuterium permeation behavior of 15-15Ti and 316L stainless steels is similar. Additionally, this study discussed the effects of microstructures and surface oxidation on the permeability of stainless steels. This study will aid in evaluating tritium permeation in lead-based fast reactors and other Gen-IV reactors and whether tritium permeation barrier or tritium-removal devices are necessary. • The deuterium transport behavior of 15-15Ti stainless steel was investigated through gas-driven permeation experiments. • The deuterium permeation behavior of 15-15Ti stainless steel is similar to that of 316L stainless steel. • The effects of microstructure and surface oxidation films on the permeability of 15-15Ti stainless steels were discussed. [ABSTRACT FROM AUTHOR]

The aspect of the detection of hydrogen isotopes in nuclear materials was analyzed in this review. Particular attention was paid to the spectroscopic methods for the detection of hydrogen isotopes in fusion reactor materials since their presence can induce severe problems in terms of material degradation and shortening of its lifetime. The main focus is on laser-induced breakdown spectroscopy, an optical emission method that is efficient, reliable, and fast. Recent results obtained at the VINCA Institute will also be presented. [ABSTRACT FROM AUTHOR]

Lithium titanate is one of the crucial breeder materials in nuclear fusion reactors. High lithium atomic density is the most desired property for this application. In view of this, advanced lithium titanate (Li2+xTiO3+y) ceramics with varying stoichiometry of Li were synthesized at temperatures in the range of 900–1100 °C and characterized using X-ray diffraction and scanning electron microscopy. External (in air) particle induced gamma-ray emission (PIGE) method was utilized for the quantification of lithium in the direct solid samples using 14N (from air) as an external current normalizer. The method was validated using certified lithium titanate reference material. The obtained results are found to be in good agreement with the theoretical values. The present study highlights the advantages of the external PIGE as a simple non-destructive method of quantification of Li in ceramics over conventional methods. [ABSTRACT FROM AUTHOR]

This paper presents the results of an experimental study of the interaction of a candidate sacrificial material (SM) for a light water reactor melt trap with corium at the Lava-B test-bench. The candidate sacrificial material is a combination of aluminum oxide and a lead layer. The idea of using such a combination of SM is based on the fact that when the lead layer interacts with corium, there will be an increase in the intensity of heat removal from the corium, as well as the chemical interaction between the corium and SM due to the high heat-conducting properties of lead. This approach will improve the efficiency of corium localization in the melt trap compared to the current set of sacrificial material. Experiments have shown active melting and boiling of lead during its interaction with corium. This is confirmed both by the readings of thermocouples and by the X-ray diffraction phase analysis of the deposit material formed on the walls of the melt receiver (MR) of the Lava-B bench, sampled after the experiment. The experiment results show that the lead layer reduces the rate of increase in the temperature of the corium and increases the rate of erosion of the ceramic part of the SM. With these circumstances, it is possible to conclude that the use of aluminum oxide with a lead layer is promising in practice. [ABSTRACT FROM AUTHOR]

Silicon carbide fibre–reinforced silicon carbide matrix (SiC f /SiC) composites are desirable in structural nuclear applications due to their low density, high temperature strength and tolerance to energetic radiation. We have subjected two industrial grades of SiC f /SiC to high energy He²⁺ ion irradiation, up to 10,000 appm implanted He at 700 °C, to determine their suitability for blanket structural applications in nuclear fusion. Minor crystallographic evolution is observed following irradiation, with evidence of phase localised stress, intragranular strain, and lattice swelling. This is attributed to vacancy production and subsequent He bubble formation, both intragranular and at grain boundaries. During post-irradiation annealing up to 1300 °C, varying degrees of He bubble evolution are observed, with the Si phase showing the highest level of instability. The fibres appear stable, with no detectable radiation-induced defects. Meanwhile, bubbles in the fibre coating and matrix grains grow and agglomerate. Despite this, no delamination or microcracking is observed. • Transmutant He production should not be critical issue for SiC fibre reinforced SiC composite structural blanket materials. • No evidence of microcracking, grain delamination or exfoliation is evident at 10,000 appm of implanted He at 700 °C. • He implantation induces <1 % volumetric swelling in β–SiC (fibre) and <0.6 % in α-SiC and Si (matrix) phases. • Point defects and He bubbles induce slight softening of SiC fibres,hardening of matrix SiC, and modulus reductionin both. • Between 800-1300 °C, He bubbles grow in all phases, with most evolution observed in the β-SiC fibre coatings. [ABSTRACT FROM AUTHOR]

The article discusses the issue of tritium production in VVER reactors in order to use it as fuel material in fusion reactors. The production of one tritium nucleus from the raw isotope 6Li requires the expenditure of one neutron. This is equivalent to neutron losses in the fuel cycle of a fission reactor. The absorption of neutrons by 6Li is partially compensated by the fuel burnup reactivity reserve in the fission reactor. However, owing to the insufficiently rapid burnup of 6Li, it acts as an absorber throughout the entire campaign. This leads to its reduction. [ABSTRACT FROM AUTHOR]

In a fusion environment, tungsten, a plasma-facing material in a reactor, is subject to the irradiation of high-energy neutrons, generating a large amount of displacement damage and transmutation products (such as rhenium, Re). We studied the evolution of defects under irradiation in W and W-Re systems using the density functional theory (DFT) and rate theory (RT) method. The results indicate that the evolution of irradiation defects is mainly affected by the irradiation dose, dose rate, and temperature. During irradiation, loops form first in W, followed by the generation of voids, which are due to the different migration energies of point defects. Higher dose rates result in a higher density and larger size of defects in tungsten. Higher temperatures cause a decrease in void density and an increase in size. The results obtained at 600 °C were in good agreement with the reported TEM data. In W-Re alloys, it is indicated that the formation of loops is delayed because Re suppresses the nucleation of loops. The dynamic introduction of Re in W stabilizes the growth of defects compared to W-Re alloys, suggesting that transmuting elements have less detrimental effects on irradiation than alloying. As defect densities and sizes were quantified under different irradiation conditions, the results provide data for the multi-scale simulation of the radiation damage and thermal/mechanical properties in plasma-facing materials under fusion conditions. [ABSTRACT FROM AUTHOR]

This document is an editorial published in the journal Frontiers in Energy Research. It discusses the importance of multi-physics and multi-scale modeling and simulation methods for nuclear reactor applications. The editorial highlights the need for comprehensive analysis of various factors such as neutronics, fuel properties, thermal and hydraulic engineering, and chemical kinetics in order to achieve safe and cost-effective operation of nuclear reactors. The editorial also mentions several research papers that have explored the coupling of different physics and scales in nuclear reactor simulations, as well as advancements in modeling and simulation tools. The document concludes by emphasizing the potential benefits of incorporating modeling and simulation in the design, operation, and maintenance of nuclear reactors. [Extracted from the article]

Ammonia (NH3) inhibition represents a major limitation to methane production during anaerobic digestion of organic material in biogas reactors. This process relies on co-operative metabolic interactions between diverse taxa at the community-scale. Despite this, most investigations have focused singularly on how methanogenic Archaea respond to NH3 stress. With a high-NH3 pre-adapted and un-adapted community, this study investigated responses to NH3 inhibition both at the community-scale and down to individual taxa. The pre-adapted community performed methanogenesis under inhibitory NH3 concentrations better than the un-adapted. While many functionally important phyla were shared between the two communities, only taxa from the pre-adapted community were robust to NH3. Functionally important phyla were mostly comprised of sensitive taxa (≥ 50%), yet all groups, including methanogens, also possessed tolerant individuals (10–50%) suggesting that potential mechanisms for tolerance are non-specific and widespread. Hidden Markov Model–based phylogenetic analysis of methanogens confirmed that NH3 tolerance was not restricted to specific taxonomic groups, even at the genus level. By reconstructing covarying growth patterns via network analyses, methanogenesis by the pre-adapted community was best explained by continued metabolic interactions (edges) between tolerant methanogens and other tolerant taxa (nodes). However, under non-inhibitory conditions, sensitive taxa re-emerged to dominate the pre-adapted community, suggesting that mechanisms of NH3 tolerance can be disadvantageous to fitness without selection pressure. This study demonstrates that methanogenesis under NH3 inhibition depends on broad-scale tolerance throughout the prokaryotic community. Mechanisms for tolerance seem widespread and non-specific, which has practical significance for the development of robust methanogenic biogas communities. Key points: • Ammonia pre-adaptation allows for better methanogenesis under inhibitory conditions. • All functionally important prokaryote phyla have some ammonia tolerant individuals. • Methanogenesis was likely dependent on interactions between tolerant individuals. [ABSTRACT FROM AUTHOR]

C276 superalloy is considered as a potential structural material for advanced nuclear reactor with good mechanical properties and corrosion resistance. High-temperature creep behaviour of C276 alloy was investigated in the temperature range of 650°C–700°C and at stresses of 140–430 MPa. A linear relationship was fitted between stress and minimum creep rate in the logarithmic coordinate system. The rupture time is analysed for life prediction in terms of isotherm extrapolation method, Monkman–Grant relation, and Larson–Miller parameter method, respectively. The isochronous stress–strain curves as a means of representing stress–strain–time relations under creep conditions were established by the parameter method. The fracture surface morphology of ruptured specimens was characterised by a scanning electron microscope to elucidate the failure mechanism. [ABSTRACT FROM AUTHOR]

508-III steel has become one of the most viable materials for heavy nuclear reactor pressure vessels due to its excellent strength and fracture toughness. However, the poor machinability of 508-III steel requires the tool to withstand extremely high mechanical stress, thermal stress, and shock during the cutting process, thereby accelerating tool failure and affecting machining efficiency. High-feed milling (HFM) is a novel rough machining technology that ensures machining efficiency and prolongs tool life. This paper aims to reveal tool failure mechanisms during HFM of 508-III steel and to indirectly evaluate the cutting performance of the HFM cutter in the context of cutting vibration and chip morphology. Dry milling experiments of 508-III steel were carried out using an HFM cutter with arc-shaped bottom edges. Experimental results show adhesion, oxidation, and diffusion are the primary tool wear mechanisms. The wear levels between each insert present significant differences, which can be explained by cutting-edge position deviations and unexpected tool movements. With the continuous accumulation of tool wear, the insert with the weakest cutting edge strength is broken, which can be attributed to the crack propagation near the cutting edge, hard spots, and the shedding of adhesive materials. When the feed per tooth is 0.8~1.0 mm, the whole milling process shows good stability, especially in the spindle direction. The chip morphology and color variations are closely related to tool wear, which can provide an essential basis for tool failure prediction. [ABSTRACT FROM AUTHOR]