Search Results (20)

Anthocyanins are important phenolic compounds in grape skins, affecting the color, oxidation resistance, and aging ability of red wine. In recent years, global warming has had a negative effect on anthocyanin biosynthesis in grape berries. Ethylene serves as a crucial phytohormone regulating the development and ripening processes of fruit; however, the specific molecular mechanism and the regulatory network between ethylene signaling and the anthocyanin biosynthesis pathway remain incompletely understood. In this study, 400 mg/L ethephon (ETH) solution was sprayed onto the surface of grape berries at the lag phase (EL-34), and the changes in anthocyanin-related genes and metabolites were explored through transcriptomic and metabolomic analysis. The results showed that ETH treatment increased Brix and pH in mature berries. In total, 35 individual anthocyanins were detected, in which 21 individual anthocyanins were enhanced by ETH treatment. However, the anthocyanin profile was not affected by exogenous ethylene. Transcriptomics analysis showed that there were a total of 825 and 1399 differentially expressed genes (DEGs) 12 h and 24 h after treatment. Moreover, key structural genes in the anthocyanin synthesis pathway were strongly induced, including VvPAL, VvCHS, VvF3H, VvF3′5′H, VvDFR and VvUFGT. At the maturity stage (EL-38), the expression levels of these genes were still higher in EHT-treated berries than in the control. ETH treatment also influenced the expression of genes related to hormone biosynthesis and signal transduction. The ethylene biosynthesis gene (VvACO), ethylene receptor genes (VvETR2, VvERS1 and VvEIN4), ABA biosynthesis gene (VvNCED2), and ABA receptor gene (VvPYL4) were up-regulated by ETH treatment, while the auxin biosynthesis gene (VvTAA3) and seven genes of the auxin-responsive protein were inhibited by exogenous ethylene. Meanwhile, ETH treatment promoted the expression of the sugar transporter gene (VvEDL16) and two sucrose synthase genes (VvSUS2 and VvSUS6). In EHT-treated berries, 19 MYB and 23 ERF genes were expressed differently compared with the control (p < 0.05). This study provides the theoretical foundation and technical support for the regulation of anthocyanin synthesis in non-climacteric fruit. Full article

This study presents a comprehensive analysis of the microstructural characteristics and chemical composition of base and weld materials from reactor pressure vessels in the first (units 1 and 2) and second (unit 8) generations of Russian VVER 440 reactors at the Greifswald nuclear power plant. We measured the specific activities of ⁶⁰Co and ¹⁴C in activated samples from units 1 and 2. ⁶⁰Co, with its shorter half-life (t_1/2 = 5.27 a), is a key dose-contributing radionuclide during decommissioning, while ¹⁴C (t_1/2 = 5700 a) plays an important role in a geological repository for low- and intermediate-level radioactive waste. Our findings reveal differences in the proportions of trace elements between the base and weld materials as well as between the two reactor generations. Microstructural analysis identified Mo-rich precipitates and (Mn, S)-rich inclusions containing secondary micro-inclusions in the unit 1 and 2 samples. Raman spectroscopy confirmed iron oxides (γ-Fe₂O₃, Fe₃O₄), silicates (Mn-SiO₃), and Cr₂O₃/NiCr₂O₄ in the base metal as well as MnFe₂O₃ in the weld metal. X-ray photoelectron spectroscopy identified Mn inclusions as MnS, MnS₂, or mixed Mn, Fe sulfides, and the Mo precipitates as MoSi₂. These findings offer valuable insights into the speciation of elements and the potential release of radionuclides through corrosion processes under repository conditions. Full article

Low carbon dioxide (CO₂) emissions make nuclear energy crucial in decarbonizing the economy. In this context, nuclear safety, and especially the operation of nuclear power plants, remains a critical issue. This article presents a new fractal cluster method of control that improves the quality of assessing fuel element cladding integrity, which is critical for nuclear and environmental safety. The proposed non-destructive testing method allows for detecting defects on the inner and outer cladding surfaces without removing the elements from the nuclear reactor, which ensures prompt response and prevention of radiation leakage. Studies have shown that the fractal dimension of the cladding surface, which varies from 2.1 to 2.5, indicates significant heterogeneity caused by mechanical damage or corrosion, which can affect its integrity. The density analysis of defect clusters allows quantifying their concentration per unit area, which is an important indicator for assessing the risks associated with the operation of nuclear facilities. The data obtained are used to assess the impact of defects on the vessel’s integrity and, in turn, on nuclear safety. The monitoring results are transmitted in real time to the operator’s automated workstation, allowing for timely decision making to prevent radioactive releases and improve environmental safety. The proposed method is a promising tool for ensuring reliable quality control of the fuel element cladding condition and improving nuclear and environmental safety. While the study is based on VVER-1000 reactor data, the flexibility of the proposed methodology suggests its potential applicability to other reactor types, opening avenues for broader implementation in diverse nuclear systems. Full article

As the cornerstone of severe accident management strategy, either in-vessel or ex-vessel retention of core melt (IVR or EVR) plays a pivotal role in the stabilization and termination of a severe accident and ultimately secures the safety goal of “Practical elimination of large radioactive release” for light water reactors. In contrast to the IVR measures that are more or less identical in reactor designs, the EVR measures are quite different from design to design. This study intended to give a critical review on the EVR measures adopted in the reactor designs of VVER-1000, EPR, ESBWR, EU-APR1400 and APWR. The review study began with a general description of the existing EVR measures, including their principles, operational procedures and research efforts. We then focused our discussions on the pros and cons of each EVR measure through the comparisons with the IVR and with the others in terms of simplicity, reliability and economy. We finally tried to identify the remaining issues and uncertainties in the qualification of the EVR measures, based on which potential design improvements and future research needs were recommended. Full article

Assessing the embrittlement and hardening of reactor pressure vessel steels is critical for the extension of the service lifetime of nuclear power plants. This paper summarises the tensile test results on the irradiation behaviour of realistic VVER-1000 welds from the STRUMAT-LTO project. The welds were irradiated at the HFR (Petten, the Netherlands) to a fluence of up to 1.087 × 10²⁰ n·cm⁻², and their irradiation hardening was studied by means of tensile testing. The four grades, with different Mn and Ni contents, show different hardening behaviours. The highest degree of irradiation hardening is observed for the weld that has the highest combined Ni + Mn content. The results show that there is a synergetic effect of Mn and Ni on the irradiation hardening behaviour of the VVER-1000 welds. Besides irradiation hardening, the effectiveness of post-irradiation annealing treatments on the recovery of the tensile properties is studied in the present work. Post-irradiation annealing treatments conducted at 418 °C and at 475 °C proved to be effective for three of the four investigated welds. For the realistic weld with the highest combined Ni + Mn, only the annealing at 475 °C led to the complete recovery of the tensile properties. Full article

It is of meaningful importance to evaluate the performance of all the nuclear facilities, and particularly those part of such buildings where spent nuclear fuel (SNF) is stored to assess what kinds of consequences are anomalous/abnormal or to determine what types of accident events may occur. In this preliminary study, the strategies adopted for the management of SNF, and the risk related to them are discussed. The aim of this study is to evaluate the total radioactivity inventory characterising Ukrainian nuclear facilities, including storage facilities. The dataset used to calculate the total activity associated with nuclear fuel is provided and discussed. For the evaluation, it is considered that a SNF pool in VVER-1000 is designed to store 687 fuel assemblies, and 670 are in VVER-440. When it is half full, which is the case for 15 Ukrainian units, it will store about 2200 tU containing up to 1·10¹⁹ Bq of 137Cs, 7·10¹⁸ Bq of 90Sr, and 1·10¹⁹ Bq of TUE. This study focuses particularly on the total activity of the SNF stored at the Zaporozhye plant, the biggest nuclear plant in Europe, and the risk posed by the potential loss that cooling the plant could incur because of pond water level variation. The results of the analysis of the Zaporozhye NPP behaviour suggest that the water flow rate which keeps the SNF pool temperature constant is about 200,000 m³·day⁻¹. Therefore, the water level in the pond should not be lower than 1.5–2 m; otherwise, the plant will need an additional source of water of more than 200,000 m³ per day to guarantee safe storage of SNF. Full article

The study considers approaches to ensuring energy management for the safe operation of facilities and their equipment and ways to improve it. It has been established that to ensure effective safety management of industrial enterprises, one of the critical areas is the technical diagnostics of power equipment during operation. An assessment of the actual technical condition of power equipment of VVER-1000 power units is proposed based on establishing the aging mechanisms and determining the relative evaluation coefficients for the characteristics of individual equipment elements. The results of the calculations allowed us to conclude that the obtained results correspond to the coefficients of relative assessment Ki of the technical characteristics of the power equipment that determine its degradation. Studies indicates that when assessing the state of power equipment, it is necessary to consider the presence and impact of the following operational factors that are not considered in the design calculations: loads, high levels of mechanical stress, fatigue damage, and metal defects, which primarily indicate the presence of degradation changes. To assess the technical condition of the equipment, considering the degree of mechanical wear, 17 technical characteristics were selected to determine the aging mechanisms by signs of degradation. A mathematical model of the dependence of the relative evaluation coefficient K on changes in the operating parameters is presented, and it is noted that the most significant influence on the value of the coefficient is the temperature of the coolant at the inlet (K = 0.56). The developed approach makes it possible to improve the safety management system of power facilities by introducing the proposed model to assess the technical conditions of power equipment by defining the parameters in the overall safety management system. Full article

Hydrogen is a flammable gas that can generate thermal and mechanical loads which could jeopardise the containment integrity upon combustion inside nuclear power plants containment. Hydrogen can be generated from various sources and disperses into the containment atmosphere, mixing with steam and air following a loss of coolant accident and its progression. Therefore, the volumetric hydrogen concentration should be examined within the containment to determine whether a flammable mixture is formed or not. Codes with 3D capabilities could serve this examination by providing detailed contours/maps of the hydrogen distribution inside containment in view of the local stratification phenomenon. In this study, a 3D VVER-1000 as-built containment model was sketched in AutoCAD and then processed into GOTHIC nuclear containment analysis code for hydrogen evaluation. The model was modified to a great extent by installing 80 passive autocatalytic recombiners and locating hydrogen sources to evaluate the performance of the hydrogen removal system inside the containment on maintaining the hydrogen concentration below the flammability limit during a large break loss of coolant accident. 2D profiles and 3D contours of volumetric hydrogen concentration with and without PARs are presented as the simulation outcome of this study. The results were validated against the results of the Final Safety Analysis Report, which also demonstrates the effectiveness of the hydrogen removal system as an engineered safety feature to keep the containment within a safe margin. Detailed 3D contours of hydrogen distribution inside containment can be employed to evaluate the local hot spots of hydrogen, rearranging and optimising the number and location of PARs to avoid the hydrogen explosion inside containment. Full article

This paper presents the results obtained from numerical evaluations for the possibility of large-scale ²³⁸Pu production in the light-water VVER-1000 reactor and the reactivity effect caused by the loss-of-coolant accident in the central fuel assembly of the reactor core. This fuel assembly containing the Np–Am-component of minor actinides was placed in the center of the reactor core and intended for intense production of ²³⁸Pu. Optimal conditions were found for large-scale production of plutonium with an isotope composition suitable for application in radioisotope thermoelectric generators. The reactivity effect from the loss-of-coolant accident in the central Np–Am fuel assembly was evaluated, and the perturbation theory was used to determine the contributions of some neutron processes (leakage, absorption, and moderation) to the total variation of the effective neutron multiplication factor. Full article

The two-phase emission detector RED-100 with 130 kg of liquid xenon as a working medium has been exhibited at a distance of 19 m from the core of the VVER-1000/320 nuclear power reactor at the fourth power unit of the Kalinin Nuclear Plant Power in 2021–2022. Due to the high sensitivity of the detector for weak ionization signals (down to single electrons), the detector has been used to search for the elastic coherent scattering of reactor electron antineutrinos off xenon nuclei. However, the observation of ~30 kHz single-electron noise did not quite allow for an effective selection of the useful events. The next experiment with the RED-100 detector is considered to be arranged with 62 kg of liquid argon as a working medium. The advantages of this approach are discussed in this paper. Full article

In this paper the phase formation and mechanical properties of VVER-type reactor pressure vessel (RPV) steels with various Ni (1.57–5.95 wt.%) and Mn (0.03–0.76 wt.%) content after neutron irradiation up to fluences in the range of (53–120) × 10²² n/m² at 400 °C were studied. The possibility of carbonitride formation under these irradiation conditions is shown. In case of sufficient Ni (>1.5 wt.%) and Mn (>0.3 wt.%) content formation of Ni-Si-Mn precipitates is observed. Their chemical composition is close to G-phase and Γ₂-phase and differs from that of radiation-induced precipitates in VVER-1000 RPV steels. This indicates the prerequisites for thermally conditioned mechanism of Ni-Si-Mn precipitates formation and growth at 400 °C enhanced by irradiation. It is also shown that the optimized steel manufacturing technology coupled with an ultralow Mn content (≤0.03 wt.%) in steel with increased up to 5.26 wt.% Ni content facilitates suppressing the Ni-Si-Mn precipitates and carbonitrides formation. This, in turn, reduces the contribution of the hardening embrittlement mechanism and, correspondingly, facilitates high radiation resistance of the steels with ultralow Mn content at the increased irradiation temperature (400 °C). Full article

Specimens of 15Kh2MFAA steel used for reactor pressure vessels V-213 (VVER-440 reactor) were studied by positron annihilation techniques in terms of their radiation resistance and structural recovery after thermal treatment. The radiation load was simulated by experimental implantation of 500 keV H⁺ ions. The maximum radiation damage of 1 DPA was obtained across a region of 3 µm. Radiation-induced defects were investigated by coincidence Doppler broadening spectroscopy and positron lifetime spectroscopy using a conventional positron source as well as a slow positron beam. All techniques registered an accumulation of small open-volume defects (mostly mono- and di-vacancies) due to the irradiation, with an increase of the defect volume ΔV_D ≈ 2.88 × 10⁻⁸ cm⁻³. Finally, the irradiated specimens were gradually annealed at temperatures from 200 to 550 °C and analyzed in detail. The best defect recovery was found at a temperature between 450 and 475 °C, but the final defect concentration of about ΔC_D = 0.34 ppm was still higher than in the as-received specimens. Full article

Laser Shock Peening (LSP) is a surface treatment technique for metallic materials. It induces plastic deformation at the surface of up to around 1 mm in depth. This process introduces residual stresses that lead to strain hardening, and potentially improvements in fatigue, stress corrosion cracking (SCC) and general corrosion behaviour in many, but not all, corrosive media. In this paper, two specimens made of AISI 304L stainless steel, one LSP-treated and one un-treated, were tested at 280 °C and 8 MPa in VVER (or PWR) primary circuit water chemistry using in situ Electrochemical Impedance Spectroscopy (EIS). This experiment serves to qualify the influence of LSP on the changes in corrosion behaviour in high-temperature, high-density water. The residual stress (RS) measurement of the surface showed a compression RS. Before LSP treatment, RS at the surface was 52.2 MPa in the rolling direction 0°RD and 10.42 MPa in the transverse rolling direction 90°RD. After the treatment, surface RS was −175.27 MPa and −183.51 MPa for Scan and TScan directions, respectively. The effect of compressive RS at the surface was studied and showed an increase in corrosion rate. The analysis of oxide layer by SEM revealed differences between LSP-treated and untreated AISI 304L specimens and their connection to corrosion rates. Full article

This paper considers the possibility for large-scale production of plutonium isotope ²³⁸Pu in the light-water nuclear power reactor VVER-1000. ²³⁸Pu is a unique source of long-term autonomous energy supply in various devices for remote regions of the Earth and in outer space. The design of the irradiation device with ²³⁷NpO₂ as a starting material is proposed, which is placed in the central zone of the VVER-1000 reactor core and makes it possible to achieve 8% of the specific Pu production (Pu/²³⁷Np) by optimizing the pitch of NpO₂-rod lattice. The computations showed that the scale of ²³⁸Pu production in the irradiation device was remarkably larger (2 ÷ 7 times more) than similar values in research reactors. At the same time, the use of heavy neutron moderators with low neutron absorption (natural lead or lead isotope ²⁰⁸Pb) around the NpO₂ fuel assembly (FA) made it possible to obtain high-purity ²³⁸Pu with the content of ²³⁶Pu below 2 ppm. The paper also shows that if the irradiation device is placed in central zone of the VVER-1000 reactor core, then the displacement damage dose in the reactor vessel remains low enough to conserve its strength properties throughout the entire period of the reactor operation (60 years). Full article

Following the recent surge in harnessing clean energy sources to fast-track carbon neutrality, renewable and nuclear energies have been the best-rated sources of clean energy. Even though renewable energy presents an almost insignificant risk to public health and the environment, they are insufficient to support the growing demand for the high energy required for industrialization. Despite the competitive potential of nuclear energy to meet these demands, public concerns about its safety have significantly hindered its mass deployment in developing countries. Therefore, one of the primary considerations in commissioning a nuclear power plant is the establishment of emergency planning zones based on the reactor type and other siting criteria. Based on Ghana’s reactor type assessment (RTA), four reactor designs were considered in this study which are APR1400, HPR1000, VVER1200, and Nuscale Power Module. Using the NRC’s SNAP/RADTRAD and RASCAL codes, this research sought to investigate radionuclide doses released at the Exclusion Area Boundary (EAB), Low Population Zone (LPZ), Control room (CR), and the 16 km recommended public safe zone during Fuel handling Accidents (FHA), Rod Ejection Accident (REA), and Long-Term Station Blackout (LTSBO). The results revealed that reactors’ power contributed to the source term activities and offsite consequences during REA and LTSBO, while FHA was predominantly affected by the number of fuel assemblies and a fraction of damaged fuel. Additionally, the accidents considered in this study followed a similar trend of impact in decreasing order of reactor power and the number of fuel assemblies; APR1400 < VVER1200 < HPR1000 < Nuscale. Nevertheless, all the doses were within regulatory limits. Full article