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A heated junction thermocouple (HJTC) is a device designed to monitor the water level of nuclear reactors to prevent explosions due to overheating. This device consists of heated and non-heated thermocouples. When heat is applied, the temperature rise in HJTC varies depending on whether the device is submerged in water or exposed to air. In this study, we analyzed the thermal characteristics of HJTC through 3D COMSOL simulation. The temperature profile of the cross-section was extracted based on three design parameters: applied heat, heating length of the wound coil, and diameter of the HJTC. The temperature difference between the two thermocouples was more pronounced upon exposure to air. Meanwhile, high applied heat, short coil heating length, and small diameter of the HJTC contributed to a more pronounced temperature difference based on the surrounding environment. These results can aid in the appropriate HJTC design to achieve the desired detection sensitivity. [ABSTRACT FROM AUTHOR]

Reactor pressure vessel (RPV) studs are key components of nuclear reactors, and their connection with flange ensures the sealing of the RPV under high-pressure and high-temperature conditions. In the present work, the external threads of the RPV stud were prepared by triaxial rolling, and the texture evolution of the external thread root material of an RPV stud was predicted by finite element analysis coupled with viscoplastic self-consistent simulations. The microstructure of the external thread root material of RPV stud was characterized by scanning electron microscope and electron back-scattered diffraction. The installation characteristics of the turned and rolled parts of the RPV stud were tested by installation and pretightening tests. It was found that the dynamic recrystallization at the external thread root formed ultrafine tempered sorbite grains, high-angle grain boundaries (47%), and strong {111} <110> and {111} <112> textures. In the installation and pretightening test, the residual elongation of rolled parts was reduced by 6% under the same loading pressure. The triaxial rolling process distributed the microstructure of the external thread root of the RPV stud in a gradient manner, resulting in improved stud installation characteristics. [ABSTRACT FROM AUTHOR]

Photodetectors used in future high-energy physics experiments will need to keep sufficient performance during a few years of data-taking despite radiation load, which, for example, in the planned upgrade of the Ring Imaging Cherenkov detectors in Large Hadron Collider beauty (LHCb) experiment is estimated at about 10 13 1-MeV neutron equivalent per cm 2 (neq/cm 2 ). Silicon photomultipliers (SiPMs) are considered as candidates for photodetectors for this application. However, their sensitivity to neutron irradiation may seriously compromise their operation, with the increase in dark count rates being the primary limitation after the irradiation. In this work, 1 × 1 mm 2 15 μ m pitch NUV-HD-RH silicon photomultipliers developed by the Fondazione Bruno Kessler were characterized before and after the irradiation. In total, 5 SiPMs were irradiated at the Jožef Stefan Institute TRIGA nuclear reactor with different fluences from 10 9 neq/cm 2 up to 10 13 neq/cm 2 . The SiPMs were also annealed at high temperatures and re-characterized after the annealing. For the SiPM characterization in all the cases, current–voltage (I-V curve) measurements, dark count rate measurements, and waveform analysis, including single photon time resolution, were carried out at different controlled temperature steps from room temperature down to the liquid nitrogen temperature. While cooling to − 20 ∘ C was enough for the SiPM irradiated at 10 9 neq/cm 2 to recover the ability to resolve single photons, by 10 13 neq/cm 2 , cooling to liquid nitrogen temperature was necessary. With sufficient cooling, the measured single photon time resolution was around 90 ps FWHM for all irradiation levels. [ABSTRACT FROM AUTHOR]

The development of reactor technologies and neutrino physics by means of nuclear reactors requires unceasingly refining and extending knowledge of reactors as an antineutrino source. By employing new data on the reactor-antineutrino spectrum and updated nuclear databases and upon taking into account special features of fuel burnup, the thermal energy released in the reactor core per fission event is calculated for a standard operating period of a thermal-neutron reactor. [ABSTRACT FROM AUTHOR]

The paper provides a detailed overview of the development of a containment pressure and temperature (P/T) analysis methodology for advanced power reactor 1400 (APR1400) nuclear power plants (NPPs). The study addresses the restrictions on exporting independent nuclear power plants by utilizing the containment analysis package (CAP) computer code developed in Korea. One of the key aspects highlighted in the paper is the comparison of results obtained from the CAP code with those from the CONTEMPT-LT/028 code which is used P/T analysis. The analysis focuses on two types of accidents: loss of coolant accidents (LOCA) and main steam line break (MSLB) accidents, which are considered as design basis accidents. By comparing the outcomes of both codes, the paper evaluates the performance and effectiveness of the CAP code in predicting the P/T behavior within the containment during these accidents. The paper also discusses the criteria and technical standards for the containment P/T analysis. It emphasizes the importance of ensuring that the peak P/T remain within the safety related systems, equipment, and structures of NPP. The design pressure is identified as a critical factor in achieving these objectives. In conclusion, the study presents the successful development of a containment P/T analysis methodology using the CAP computer code for APR1400. The methodology considers the specific characteristics of Korean NPPs and new code, CAP. The paper emphasizes the applicability and effectiveness of the CAP code in this context. However, further research and validation efforts are recommended to enhance the accuracy and reliability of the methodology for various design basis accidents. The developed methodology is expected to contribute to the safe and efficient operation of APR1400 NPPs and support Korea's ambitions in exporting NPPs' technology to other countries. [ABSTRACT FROM AUTHOR]

Accident tolerant fuel (ATF) cladding is a new type of nuclear fuel cladding designed to improve the safety and performance of nuclear reactors. In this paper, the kinetics and degradation mechanisms during high-temperature oxidation in steam of the three most promising ATF cladding materials, i.e., chromium-coated zirconium alloys, FeCrAl alloys, and silicon carbide-based composites, are described. Each system has its own degradation mechanisms leading to different maximum survival temperatures. After providing general information and data to understand the oxidation and degradation processes, illustrative examples obtained at the Karlsruhe Institute of Technology are given for each type of cladding. The maximum temperatures at which the barrier effect of the cladding can be maintained for a reasonable period of time during nuclear accident scenarios are 1200–1300 °C for Cr-coated Zr alloys, 1400 °C for FeCrAl alloys, and 1700 °C for SiC-based composite claddings. [ABSTRACT FROM AUTHOR]

Microstructure-aware models are necessary to predict the behavior of material based on process knowledge or to extrapolate mechanical properties of materials to environmental conditions which are not easily reproduced in the laboratory, e.g., nuclear reactor environments. Elemental Ta provides a relatively simple BCC system in which to develop a microstructural understanding of deformation processes which can then be applied to more complicated BCC alloys. In situ neutron diffraction during compressive deformation and subsequent heat treatment have been used to monitor the evolution of microstructural features in Ta throughout simulated processing steps. Crystallographic texture and dislocation density are determined as a function of first plastic strain, then temperature. Lattice strains are determined and attributed to stresses at macroscopic, grain and dislocation length scales. The increase of the dislocation density through deformation and subsequent recovery during heat treatment is monitored through the changing diffraction line profile. Also, randomization of the texture is used as a signature of recrystallization. The recovery of dislocations through annihilation is not observed to depend on the initial dislocation density in the range studied here. In contrast, recrystallization is observed to depend strongly on the initially dislocation density. [ABSTRACT FROM AUTHOR]

SA508Gr.3Cl.2 steel is extensively employed in fabricating nuclear reactor facilities consisting of reactor pressure vessels and steam generators. The functional performance of components is affected by the properties of superficial layers, and the machining process is indispensable for the final forming. Hence, it is essential to modify and improve surface quality via the regulation of machining parameters. This paper presents a comprehensive investigation of the effects of high-speed face milling process on the surface modification of SA508Gr.3Cl.2 steel, with particular attention to surface topography, microstructure alteration, micro- and nano-hardness, and residual stresses under semi-finishing conditions. To control the machined surface integrity via regulating milling parameters, surface characteristics are compared under various milling speeds and feeds. Results indicate that the unwanted geometrical defects in the form of feed marks, plowing grooves, surface tearing, and side flow are inherited on the machined surface, and the surface roughness is maintained at a relatively lower level, up to Sa 0.822 μm and Sz 5.824 μm. In addition, the material dragging layers with different thicknesses between 4.1 and 8.5 μm are introduced into the regions near the machined surface, which are subjected to intense plastic activities, including dislocation proliferation, grain boundary deflection, grain refinement, and elongation. There is no evidence of white layer and phase transformation at the superficial layer of machined SA508Gr.3Cl.2 steel. Besides, the increased milling speed and feed are responsible for thicker plastic deformation layers and greater hardening depths and promote surface hardening. A cutting configuration with a higher milling speed and lower feed is recommended to obtain attractive component surfaces with fewer geometrical defects, lower roughness, lower tensile residual stresses, proper hardening, and no white layer and phase transformation. [ABSTRACT FROM AUTHOR]

Sensor degradation poses a critical yet 'often overlooked' challenge in accurately predicting the remaining useful life (RUL) of nuclear reactor pressure vessels (RPVs), hindering safe and efficient plant operation. This paper introduces an approach to RUL estimation that explicitly addresses sensor degradation, a significant departure from conventional methods. We model neutron embrittlement, a dominant degradation process in RPV steel, as a Wiener process and leverage real-world surveillance capsule data for insightful parameterization. Maximum likelihood estimation is utilized to characterize the degradation dynamics in the model. A Kalman filter then seamlessly integrates the degradation model with sensor measurements, effectively compensating for degradation-induced errors and providing refined state estimates. These estimates power a robust RUL prediction framework. Our results expose the profound impact of sensor degradation on conventional RUL predictions. By directly confronting sensor degradation, our method yields substantially more accurate and reliable RUL estimates. This work marks a significant advancement in the field of materials degradation, offering a powerful tool to optimize nuclear power plant safety and longevity. [ABSTRACT FROM AUTHOR]

Several neutron sources are available at Research Centre Řež (CVŘ) that are usable in neutron physics and radioanalytical measurements, namely two experimental nuclear reactors—a 'zero' power (maximum 5 kW) LR-0 reactor and a medium power (maximum 10 MW) LVR-15 reactor, a 252Cf source and a D–T generator of fast neutrons. Their basic parameters and modes of applications in various fields of science and technology are described. The reactors and other neutron sources are equipped with a number of gamma-ray (mostly High Purity Germanium (HPGe)), and neutron spectrometers to allow for assay of studied materials. Most of the above facilities are available in an open-access regime. [ABSTRACT FROM AUTHOR]

An obsolete control unit of a fast transport (rabbit) system for short-time neutron activation analysis at the LVR-15 experimental reactor at Řež has been replaced. In the new system a PC is used to fully control the rabbit system and the subsequent gamma-ray spectrometry measurement. The modernization resulted in significant simplification of the operator work and improved documentation of the irradiation and spectra acquisition procedures. New nuclear safety features of the rabbit system are also described. [ABSTRACT FROM AUTHOR]

The heavy water reactor nuclear fuel channel is loaded with nuclear fuel bundles, therefore, its structural integrity is important. Since the fuel channel structure is affected by radiation irradiation as well as flow-induced vibration due to heavy water coolant flow, it is necessary to monitor and diagnose its structural health condition in real time. Therefore, measurement systems of neutron noise and structural vibration are utilized in heavy water reactor system. In this paper, we introduce a research regarding an AI (artificial intelligence) system that can monitor the presence of mechanical defects at supports of the nuclear fuel channel, and accurately locate the failure area using vibration data of the nuclear fuel channel and CNN (convolution neural network). Vibration data of the fuel channel for normal and abnormal conditions were generated by fluid-structure interaction analysis. CNN and AI systems were trained to recognize different vibration characteristics of normal and various failure modes using images of short time Fourier transform. It was confirmed that the proposed AI accurately identifies the presence of mechanical defect of support structures and its location, even when the mass flow rate of heavy water varies. The results of this research can also be applied to develop AI system for fault diagnosis of nuclear fuel channels based on neutron noise data obtained by neutron flux detectors installed in the actual heavy water reactor so that the structural integrity of the heavy-water nuclear power plant system could be ensured more reliably. [ABSTRACT FROM AUTHOR]

The energy crisis and climate change have promoted a growing interest in non-fossil sources, such as nuclear, with uranium carbides being seen as potential fuel candidates for Generation IV nuclear reactors. However, the need of accurate thermophysical data for the fuel and its compatibility with core materials during the extreme fission conditions is still an issue. Here a study of the ternary uranium-iron-carbon system performed at 1100 °C using powder x-ray diffraction and Scanning Electron Microscopy coupled with Energy Dispersive Spectrometry is presented. The U-Fe-C isothermal section is characterized by two ternary compounds, thirteen 3-phase regions and five 2-phase regions. UFeC2 and ~ U11Fe12C18 were confirmed to be present at 1100 °C and crystallize in structures related to the binary uranium carbides. UFeC2 crystallizes in an original structure type, a distorted variant of the UCoC2 structure, with space group P4/n and a = 3.503(5) Å and c = 7.405(5) Å lattice parameters. ~ U11Fe12C18, has a crystal structure related to the Th11Ru12C18 structure-type (space group I 4 ¯ 3m) with the lattice parameter a ≈ 10 Å. Furthermore, an island of a α-UC2-based phase with 32U:4Fe:64C composition was found in the 1100 °C isothermal section, indicating the inclusion of Fe in the α-UC2 binary compound. [ABSTRACT FROM AUTHOR]

The paper discusses a photoactivation method for determining the activity of the long-lived isotope Ca relative to the activity of Co in irradiated reactor biological shielding concrete. It is proposed to use photonuclear reactions on cobalt and calcium isotopes: Co( , ) Co, Ca( , ) K, or Ca( , ) Ca Sc to determine the activity of Ca. The sensitivity of the proposed method is on the order of Bq/g, provided that semiconductor spectrometers with an ultrapure germanium detector are used. The structure and chemical compositions of the sample do not affect the proposed method. It avoids the long sample preparation times associated with radiochemical methods. [ABSTRACT FROM AUTHOR]

This study deals with the stochastic Fitz-Hugh Nagumo (FHN) equation and its multiple soliton solutions. The underlying model has numerous applications in neuroscience that express the pulse behavior of neurons. In general, different kinds of noise affect neurons, e.g., oscillations in the opening and closing of ion stations within cell membranes and the fluctuations of the different conductivities in the system. This fluctuation creates a sequence of stochastic excitation. Various applications are branching Brownian motion process, flame propagation, nuclear reactor theory, autocatalytic chemical reaction, mobility in neurons, population growth in the open environment, and liquid environment. So, need of the hour to consider the FHN equation under the impact of noise. The ϕ 6 -model expansion method is used to extract the analytical solutions that give a dynamic attitude of the transmission for the nerve impulses of a nervous system. The different constraint conditions for the existence of these solutions are also discussed. The solutions of this model are represented in hyperbolic, trigonometric, and rational forms. The 2 and 3-dimensional behavior of these solutions are depicted by choosing the different values of parameters. The impact of noise on the physical system is analyzed and its real-world applications are discussed. The spikes in the solutions are controlled through the Borel function. These important results will open a new horizon of research for the young researchers. [ABSTRACT FROM AUTHOR]

Incidents involving chemical storage tanks in the petrochemical industry are significant events with severe consequences. Within the petrochemical industry, EDC is a sector that produces ethylene dichloride through the reaction of chlorine and ethylene. The present research was conducted to evaluate the consequences of chlorine gas released from the EDC reactor in a petrochemical industry in southern Iran. Data regarding reactor specifications were obtained from the factory's technical office, while climatic data was acquired from the Meteorological Organization. The consequences of chlorine gas release from the reactor were assessed in four predefined scenarios using numerical calculation methods and modeling with the ALOHA software. The numerical calculation method involved thermodynamic fluid path analysis, discharge coefficient calculations, and wind speed impact analysis. The hazard radius was determined based on the ERPG1-2–3 index. Results showed that in the scenario of chlorine gas release from EDC reactors, according to the ALOHA model, an increase in wind speed from 3 to 7 m/h led to an expanded dispersion radius. At a radius of 700 m from the reactor, the maximum outdoor concentration reached 3.12 ppm, decreasing to 2.27 ppm at 800 m and further to 1.53 ppm at 1000 m. The comparison of numerical calculations and modeling using the ALOHA software indicates the desirable conformity of the results with each other. The R2 coefficient for evaluating the conformity of the results was 0.9964, indicating the desired efficiency of the model in evaluating the consequences of the release of toxic gasses from the EDC tank. The results of this research can be useful in designing the site and emergency response plan. [ABSTRACT FROM AUTHOR]

Homogeneous aqueous solution nuclear reactors have many advantages in the production of medical isotopes. However, the fuel precipitation problem must be addressed in transitioning the technology from research to a commercial industrial environment. The turbidity value can indirectly reflect the amount of precipitation in the fuel solution. This article systematically measured the turbidity values of fuel solutions under different nitric acid concentrations, temperatures and uranium concentrations, and studied their effects on the precipitate formation of uranyl nitrate fuel, provided some help for understanding the fuel properties of homogeneous aqueous solution nuclear reactor. [ABSTRACT FROM AUTHOR]

Nuclear reactors will face the problem of decommissioning at the end of their operating life due to the high radioactivity of reactor components and environmental safety considerations. The Heavy Water Research Reactor (HWRR) is the first large-scale research reactor to be decommissioned in China. The second phase of HWRR decommissioning involves the main components in the reactor block, so the radiation source terms and the radioactive waste level need to be evaluated before demolition and disposal. Based on the operating history, three-dimensional geometry, materials, and other information of the HWRR, the activity of radionuclides in the main components of HWRR is calculated and analyzed, and the MCNP/ORIGEN coupling scheme is utilized for theoretical analysis. The theoretical results indicate that 14C, 54Mn, 55Fe, 60Co, 63Ni, and 152Eu are the main radioactive nuclides. The total activity of radioactive nuclides was 2.36E + 15 Bq at the end of 2007, 4.27E + 13 Bq at the end of 2021, and 1.83E + 13 Bq at the end of 2025. Furthermore, local sampling and radiometric analyses based on the HPGe gamma-ray spectrometer are also performed to verify the theoretical results, the ratio of theoretical activity values to the measured activity of the experimental sample is within 2.5 times, so the theoretical results are conservative. According to the classification standards for radioactive waste, the inner shell, outer shell, cooling water tank, sand layer, and heavy concrete shielding layer are all low-level waste. These results and conclusions can serve as a reference for the second phase decommissioning of the HWRR and the subsequent disposal of radioactive waste. [ABSTRACT FROM AUTHOR]

Purpose: Classical brachytherapy of solid malignant tumors is an invasive procedure which often results in an uneven dose distribution, while requiring surgical removal of sealed radioactive seed sources after a certain period of time. To circumvent these issues, we report the synthesis of intrinsically radiolabeled and gum Arabic glycoprotein functionalized [169Yb]Yb2O3 nanoseeds as a novel nanoscale brachytherapy agent, which could directly be administered via intratumoral injection for tumor therapy. Methods: 169Yb (T½ = 32 days) was produced by neutron irradiation of enriched (15.2% in 168Yb) Yb2O3 target in a nuclear reactor, radiochemically converted to [169Yb]YbCl3 and used for nanoparticle (NP) synthesis. Intrinsically radiolabeled NP were synthesized by controlled hydrolysis of Yb3+ ions in gum Arabic glycoprotein medium. In vivo SPECT/CT imaging, autoradiography, and biodistribution studies were performed after intratumoral injection of radiolabeled NP in B16F10 tumor bearing C57BL/6 mice. Systematic tumor regression studies and histopathological analyses were performed to demonstrate therapeutic efficacy in the same mice model. Results: The nanoformulation was a clear solution having high colloidal and radiochemical stability. Uniform distribution and retention of the radiolabeled nanoformulation in the tumor mass were observed via SPECT/CT imaging and autoradiography studies. In a tumor regression study, tumor growth was significantly arrested with different doses of radiolabeled NP compared to the control and the best treatment effect was observed with ~ 27.8 MBq dose. In histopathological analysis, loss of mitotic cells was apparent in tumor tissue of treated groups, whereas no significant damage in kidney, lungs, and liver tissue morphology was observed. Conclusions: These results hold promise for nanoscale brachytherapy to become a clinically practical treatment modality for unresectable solid cancers. [ABSTRACT FROM AUTHOR]

This article presents a novel a novel incremental–iterative technique in solving the nonlinear static response of bi-directional functionally graded (BDFG) porous plate including the neutral surface effect and embed in elastic medium for the first time. A four variables high-shear deformation theory without shear correction factors is exploited to portray the kinematic relations with respect to neutral surface rather than mid-plane. BDFG material distributions are described by power functions through both thickness (z-direction) and axial (x-direction), and internal voids and porosities are included by different cosine functions. The large deformations are included within the sense of nonlinear von Kármán strains. Winkler–Pasternak type model is used to describe the elastic foundation of structures. The integro-differential equilibrium equations with associated modified boundary conditions are solved numerically and iteratively by using 2D differential/integral quadrature method (DIQM). Model validations and parametric analysis are depicted to present the influence of neutral axis, nonlinear strains, gradation indices, elastic foundations, and modified boundary conditions on the static deflection in addition to normal and shear stresses. The proposed model can be implemented in designing the static behavior and stress analyses of the nuclear reactors, marine and aerospace structures under nonlinear and large deformations. [ABSTRACT FROM AUTHOR]

A new non-destructive method has been proposed to determine the radioactivity of long-lived nickel radioisotopes produced in nuclear reactor metallic structural materials. We offer to use an electron accelerator and photonuclear reactions to determine the activity of 63Ni in spent reactor steel and alloys by the 60Co activity. The relative activity measurements allow to ignore factors such as material heterogeneity, differences in sample density and chemical composition. Using a semiconductor spectrometer with an HPGe detector, the developed method has an uncertainty of 5–10% and a sensitivity of 0.5 Bq g–1. It makes it possible to simplify the identification and control of 63Ni in metallic structural materials of nuclear power units and various types of radioactive waste. The proposed method was tested on RBMK-1000 and WWER-1000 reactor samples. [ABSTRACT FROM AUTHOR]

The Anaerobic Baffled Reactor (ABR) is an effective solution for landfill leachate treatment using an anaerobic fermentation process, which helps to reduce operating costs and sludge volume. To better understand the biological, chemical, and physical processes involved, especially when combining the ABR with an aerobic component, the study aimed to investigate the performance of an Anaerobic–Aerobic Hybrid Baffled Reactor (AABR) that includes an Anaerobic Filter (AF) for treating landfill leachate. This research utilized two glass reactors. The first reactor, designated as AABR-AF, consisted of six independent rectangular glass chambers arranged side by side. The third and sixth chamber designed for aerobic treatment and AF, respectively. The second reactor was used as a control reactor and did not include any aerobic chamber. The highest Removal Efficiencies (REs) for turbidity, COD, BOD, TP, TKN, nitrate, TOC, and TSS in the AABR-AF and ABR-AF were found to be (65.4% and 56.3%), (98.3% and 94.1%), (98.1% and 93.2%), (86.4% and 65%), (89.2% and 76.7%), (81.2% and 64.4%), (88.2% and 79.4%), and (72.4% and 68.5%), respectively. These optimal REs were achieved at an HRT of 48 h and an OLR of 10 kg/m3.d. Also, the highest and the lowest REs in Heavy Metals (HMs) were 89.57% for manganese in AABR-AF and 6.59% for nickel in ABR-AF, in an OLR of 10 kg/m3.d, respectively. The effective removal of Organic Matters (OMs) from landfill leachate using the AABR-AF and ABR-AF was found to be strongly influenced by HRT and OLR. The AABR-AF configuration, featuring a single aerobic chamber in the reactor, exhibited a higher efficiency in removing OMs compared to the ABR-AF configuration. [ABSTRACT FROM AUTHOR]

Enhancing fatigue life of thick-walled cylinders used for high-pressure application is an important design task. Cylindrical vessels find applications in oil and gas industries, food and chemical industries, nuclear reactors and gun barrels. They are susceptible to cracking due to high tensile stress caused by internal pressurization at their inner side during operation. Compressive residual stress can potentially help in suppressing inner cracks in such components to safeguard against any awkward situations. In this work, a hybrid rotational-swage autofrettage method is proposed to induce beneficial compressive residual stresses at the interior wall of a thick-walled cylinder. An analytical model is developed to determine the residual stresses in the process using Tresca yield criterion and its associated flow rule for non-hardening material behavior. The compressive residual stress generated in the hybrid autofrettage process is assessed for increase in the fatigue life of pressurized cylinders. To exemplify the process, cylinders of two different materials, viz. SS316 and Al7075-T6, are considered to be rotated while interfering at their inner side with an axially driven oversized AISI4340 mandrel. The design speeds of the cylinders for each material are obtained as a function of the swaging mandrel interference without contact separation during the process. The residual stresses in the cylinders are evaluated numerically, and the performance of the cylinders on enhancing the fatigue life under in-service internal pressurization is investigated using Paris law. [ABSTRACT FROM AUTHOR]

Autonomous nuclear reactor monitoring is a key aspect of the International Atomic Energy Agency's strategy to ensure nonproliferation treaty compliance. From the rise of small modular reactor technology, decentralized nuclear reactor fleets may strain the capacities of such monitoring and requires new approaches. We demonstrate the superior capabilities of a gamma detection system to monitor the criticality of a zero power nuclear reactor from beyond typical vessel boundaries, offering a powerful alternative to neutron-based systems by providing direct information on fission chain propagation. Using the case example of the research reactor CROCUS, we demonstrate how two bismuth germanate scintillators placed outside the reactor vessel can precisely observe reactor criticality using so called noise methods and provide core status information in seconds. Our results indicate a wide range of applications due to the newly gained geometric flexibility that could find use in fields beyond nuclear safety. [ABSTRACT FROM AUTHOR]

The article is devoted to the analysis of possible spatiotemporal kinetic structures that can arise during catalytic oxidation reactions on metal surfaces at atmospheric pressure. The catalytic oscillatory reaction in a flow reactor is modeled using a 1D system of equations of the reaction–diffusion–convection type. The STM type oscillatory reaction model of catalytic oxidation is used as a kinetic model. The obtained results of mathematical modeling show the decisive influence of an axial mixing in the reactor on the development of spatiotemporal structures. It is also shown that, depending on the ratio of adsorption rates of reacting species, three different isothermal spatiotemporal structures can arise, namely a spatially inhomogeneous stationary state, regular and aperiodic "breathing structures". [ABSTRACT FROM AUTHOR]

The scenarios of deposition on the sea surface around the Korean Peninsula following hypothetical release at Kori NPP were studied by application of the atmospheric transport model FLEXPART. The series of 1460 dispersion scenarios was calculated, covering the time period of 1 year. The total depositions on the surfaces of the East/Japan Sea (EJS), Yellow Sea (YS), East China Sea (ECS), and near-coast zone of 200-km vicinity around the Korean Peninsula, normalized on the total emission inventory were analyzed. The normalized deposition on the total area of 3 seas varied by more than 2 orders of magnitude from 0.003 to 0.59. The fraction of wet deposition in total deposition varied from 0 to 0.99. The selected most heavy deposition scenarios were divided into those dominated by dry deposition and wet deposition. The most conservative release scenarios (7 August 2020 and 2 September 2020) were dominated by wet deposition and were caused by the monsoon rainband and passage of the typhoon. In the scenario of 7 August 2020, 96% of deposition appeared in the near-coast zone. Cluster analysis of dispersion scenarios confirmed strong seasonal dependence of deposition patterns. For EJS and YS, the maximum average deposition was reached in the 'Summer' cluster. For ECS the maximum average normalized deposition was reached in the 'Spring + Fall' cluster ( D r = 0.055 ). Minimum average deposition on all seas was obtained for the 'Winter' cluster. [ABSTRACT FROM AUTHOR]

This study examines an important aspect of energy history and policy: the intertwinement of energy technologies with ideals. Ideals play an important role in energy visions and innovation pathways. Aspirations to realize technical, social, and political ideals indicate a long-term commitment in the design of energy systems, distinguishable from commitment to other abstract goals, such as values. This study offers an analytical scheme that could help to conceptualize these differences and their impact on energy policy. In the proposed model, two spheres of interaction are highlighted: a material sphere in which values and technologies co-evolve, and an imaginary sphere in which ideals interact with idealized technologies. Furthermore, the relation between these two spheres can be understood in different ways. This study examines three cases that are illustrative of the different roles of ideals in the development of energy technologies and visions: (1) the evolution of safety in nuclear reactor design; (2) visions of atomic power in France; (3) the political idealization of a tidal power scheme in the Severn Estuary. Finally, the developed model implies more general insights for the development of sociotechnical systems. Amongst others, it shows why certain projects and technologies remain a political, but not a techno–economic option. [ABSTRACT FROM AUTHOR]

The unwarranted leakage/release of hydrogen gas from metal processing, automotive, petrochemical industries, and nuclear reactors, along with its subsequent ignition and transition to detonation, could lead to catastrophic damage to both life and property. The development of practical hazard prevention and safety control systems demands an understanding of the effectiveness of the chemical inhibitors to suppress/mitigate a detonation wave under varying operational conditions. In the current study, the inhibition efficiency of chemical inhibitors under varying mixture initial conditions was investigated using numerical computations. The inhibition efficiency of trifluoroiodomethane (CF 3 I), carbon dioxide (CO 2) , and steam (H 2 O) on hydrogen-oxygen/air mixtures was evaluated using a detailed chemical kinetic model for hydrogen oxidation. ZND computations were carried out over a range of initial mixture composition, pressure, and temperature. It was found that CF 3 I is a better inhibitor than CO 2 and H 2 O at all the initial mixture conditions. However, at very high temperatures, the inhibitors CF 3 I, CO 2 , and H 2 O have a similar detonation inhibition efficiency. The inhibition efficiency of carbon dioxide and steam is comparable and significantly lower than CF 3 I. The findings from the current work can be used to design optimized detonation safety systems over a range of practical operating conditions. [ABSTRACT FROM AUTHOR]

The problem of finding the natural frequencies and eigenfunctions of the fuel element of a nuclear reactor is considered. A rod with a piecewise constant material density in the longitudinal direction is used as a model. The frequency equation is explicitly derived, and the natural frequencies of the first twelve harmonics are determined numerically and analytically. Analytical expressions for eigenfunctions are obtained, which are in good agreement with calculations made using the Galerkin method and numerical calculations in the ANSYS program. The results will be used to simulate the dynamics of the NEPTUN pulsed reactor for examining the stability of its operation. [ABSTRACT FROM AUTHOR]

The present study explores the status of nuclear energy developments for generation of electric power. Notably, India's pivotal role in advancing nuclear technology is highlighted, particularly with the development of a cutting-edge nuclear Fast Breeder Reactor (FBR) technology-based power plant. India has been developing the latest nuclear Fast Breeder Reactor (FBR) technology-based power plant having power generation capacity 10 times higher than the existing nuclear power plants. The review incorporates the most important component of the FBR system i.e. cold trap. Model Sodium Cold Trap (MSCT) is used by India to demonstrate the regeneration of secondary sodium cold trap.The development, advancements and challenges towards the Model Sodium Cold Trap (MSCT) are discussed in detail. It is possible to increase the maximum in-situ regeneration of MSCT up to 82%. This paper also reviews status of energy production using renewable and non-renewable energy sources. Finally, the scope in the advancements of nuclear reactor technology and related future opportunities has also been included. [ABSTRACT FROM AUTHOR]

This study introduces a novel method for enhancing Boiling Water Reactor (BWR) operation simulations by integrating machine learning (ML) models with conventional simulation techniques. The ML model is trained to identify and correct errors in low-fidelity simulation outputs, traditionally derived from core physics computations. These corrections aim to align the low-fidelity results closely with high-fidelity data. Precise predictions of nuclear reactor parameters like core eigenvalue and power distribution are crucial for efficient fuel management and adherence to technical specifications. Current high-fidelity transport calculations, while accurate, are impractical for real-time predictions due to extensive computational demands. Our approach, therefore, utilizes the standard two-step simulation process-assembly-level lattice physics calculations followed by whole-core nodal diffusion computations-to generate initial results, which are then refined using the ML-based error correction model. The methodology focuses on improving simulation accuracy in regular BWR operations rather than developing a universal ML predictor for reactor physics. By training an advanced neural network model on the difference in high-fidelity and low-fidelity simulations, the model can reduce the nodal power error from low-fidelity simulations to around 1% on average and the core eigenvalue down to under 100 pcm. This result is under the condition of the normal variations of control rod pattern and core flow rate changes in standard BWR operations used in the training and evaluation of the machine learning model. This work suggests a promising approach for achieving more accurate, computationally feasible simulation solutions in nuclear reactor operation and management. [ABSTRACT FROM AUTHOR]

This paper proposed a modeling method of the thermo-fluid-solid coupling simulation for the hot water pipeline with the insulation layer outside in the nuclear reactor. The temperature and stress distributions are analyzed and compared with the tests; the effects of the insulation layer thickness and support interval are investigated. The maximum stress occurs at the support points. The maximum stress increases approximately linearly with the increase of the support interval. The maximum deflection occurs at the middle of the two supports. The maximum deflection increases rapidly with the increase of the support interval, which is approximately in a parabolic relationship. The axial elongation ratio increases with the insulation layer thickness. The simulation results agree with the test results very well with regard of both stress and temperature distributions. The simulation method can also be used for the stress and temperature analysis in various cases, such as chemical-vessel and gas-oil pipeline with various thermal fluids inside. [ABSTRACT FROM AUTHOR]

In this article, we present a straightforward in-situ approach for producing Ag NPs incorporated in graphene oxide (GO) blended with glutaraldehyde (GA) cross-linked polyvinyl alcohol (PVA) matrix. Samples are γ-irradiated by doses of 2, 5, and 10 kGy and in comparison with the pristine films, the thermal conductivity ('k') and effusivity are measured. 'k' decreases with irradiation doses up to 5 kGy and further increase in the dosage results increase in 'k'. We performed FDTD modeling to verify the effect of polarization and periodicity on the absorptivity and emissivity spectra that are correlated to the 'k' and effusivity, empirically. Hence, we can confess that the structural properties of the prepared hybrid nanocomposite are manipulated by γ-irradiation. This attests that the PVA/GO-Ag/GA nanocomposite is radiation-sensitive and could be employed for thermal management systems. Moreover, their strong electrical insulation, as the measured dc conductivity of the γ-irradiated samples is found to be in the range of 2.66 × 10−8–4.319 × 10−7 Sm−1, which is below the percolation threshold of 1.0 × 10−6 Sm−1, demonstrates that they are excellent candidates for the use of thermal management materials. The low 'k' values allow us to use this promising material as thermal insulating substrates in microsensors and microsystems. They are also great choices for usage as wire and cable insulation in nuclear reactors due to their superior electrical insulation. [ABSTRACT FROM AUTHOR]

Using fuel campaign data from the third unit of the Kalinin Nuclear Power Plant, it has been shown that the inclusion of the detailed geometry of the reactor, its finite dimensions, and the distributions of the energy release and the fuel composition in the reactor core results in the appearance of inhomogeneity in the calculated antineutrino flux from it. This inhomogeneity is particularly pronounced at distances comparable with the dimensions of the reactor core. The detailed calculation gives a higher antineutrino flux at such distances than simplified models and indicates the directional dependence of the flux. This can noticeably affect measurements of the oscillation parameters and for neutrino detectors used to search for transitions of neutrinos to a sterile state at small distances from the reactor. It has been shown that not only a change in the isotopic composition of the fuel but also the variation of the energy release distribution in the reactor core can be evaluated from the variation of the count rate of a neutrino detector in specific directions to the reactor. [ABSTRACT FROM AUTHOR]

In this paper, we describe an activity that involves economics and mathematics. It is included in the planning of orientation paths towards university studies within the Mathematical High School Project and is dedicated to students in the last years of high school. In particular, this research will deal with the issue of solving an economic problem using not only real analysis instruments but also geometrical topics concerning Euclidean geometry and topology. Mathematics becomes a language to understand and explain a real life problem, such as determining the optimal position of an airport, a nuclear reactor and so on. Some activities made use of dynamic geometry software and computer simulations. [ABSTRACT FROM AUTHOR]

The annual consumption of tritium in a fusion reactor is estimated at ~56 kg per 1 GW. Its amount depends on the power of the reactor, plasma parameters, and the technologies used. Tritium can be produced in nuclear reactors to form an initial margin. The present article analyzes the demand for tritium during the future development of fusion energy. It is shown that tritium demand under various scenarios exceeds current production capabilities and margins. An overview of tritium production in PWRs (USA) is presented. Lithium absorbing rods and stages of their technological implementation are described. The production of tritium in a reactor limited to 1.5 kg/year is demonstrated to cause no effect on operation or safety. The introduction of such technologies, which have been worked out over a period of around 15 years, is possible in both VVER and RBMK reactors. [ABSTRACT FROM AUTHOR]

The results of experimental studies of the hydrodynamics of a coolant in the outlet section of a cassette fuel assembly of the RITM reactor of a low-power nuclear power plant are presented. The purpose of the work was to study the redistribution of axial velocity and coolant flow rate at the outlet of the fuel rod bundle, behind the fuel assembly cap, near the coolant extraction pipe, and in the holes of the upper support plate. To achieve the goal, experiments were carried out on a research stand with an air working medium on a model of the outlet section of the fuel assembly, including the outlet fragment of the fuel rod bundle, the cap, the upper support plate with the coolant extraction pipe. When conducting research, the pneumometric method and the method of injection of a contrast impurity were used. An area covering the entire cross section of the model was chosen as the study area. The picture of the coolant flow is represented by cartograms of the distribution of axial velocity and coolant flow rate, as well as cartograms of the distribution of the contrast impurity. The experimental results can be used in the design of new active sections of RITM reactors. The resulting experimental base can be used for local validation of CFD programs and onedimensional thermal-hydraulic codes used to substantiate the thermal reliability of cores. [ABSTRACT FROM AUTHOR]

The cross section of the 89 Y(n, γ ) reaction has important implications in nuclear astrophysics and for advanced nuclear technology. Given its neutron magic number N = 50 and a consequent small neutron capture cross section, 89 Y represents one of the key nuclides for the stellar s-process. It acts as a bottleneck in the neutron capture chain between the Fe seed and the heavier elements. Moreover, it is located at the overlapping region, where both the weak and main s-process components take place. 89 Y, the only stable yttrium isotope, is also used in innovative nuclear reactors. Neutron capture and transmission measurements were performed at the time-of-flight facilities n_TOF at CERN and GELINA at JRC-Geel. Resonance parameters of individual resonances were extracted from a resonance analysis of the experimental transmission and capture yields, up to a neutron incident energy of 95 keV. Even though a comparison with results reported in the literature shows differences in resonance parameters, the present data are consistent with the Maxwellian averaged cross section suggested by the astrophysical database KADoNiS. [ABSTRACT FROM AUTHOR]

This article presents two newly developed radiochemical procedures for the destructive analysis of niobium plates irradiated in the core of water-water energetic nuclear reactor (WWER), as well as determination of the specific activity of the generated during the irradiation metastable isotope 93mNb. The procedures were developed and tested using standard samples containing 93mNb and 94Nb. Measurements were made by low energy gamma spectrometry (LEGS) and liquid scintillation counting (LSC) and the obtained data were verified by interlaboratory comparison. All the results were compared with the calculated ones to confirm and evaluate the accuracy and superiority of the two measurement methods. [ABSTRACT FROM AUTHOR]

Characterisation of activated reactor components is a multifaceted effort, which include both experimental work and activation calculations which together form a comprehensive characterisation approach. However, both the experimental work and activation calculations have their constraints and uncertainties. This work presents the characterisation of a highly activated stainless steel pipe including sampling planning, sampling, sub-sampling, elemental and radionuclide analyses, and activation calculations. 55Fe, 63Ni and 60Co analyses were carried out in a trilateral intercomparison exercise and the results were compared with activation calculations. The results showed excellent alignment between experimental and activation calculation results. [ABSTRACT FROM AUTHOR]

The reasonable geometry design of non-thermal plasma (NTP) reactor is significant for its performance. However, optimizing the reactor structure has received insufficient attention in the studies on removing volatile organic compounds by NTP. Several dielectric barrier discharge (DBD) reactors with various barrier thicknesses and discharge gaps were designed, and their discharge characteristics and toluene degradation performance were explored comprehensively. The number and intensity of current pulses, discharge power, emission spectrum intensity and gas temperature of the DBD reactors increased as barrier thickness decreased. The toluene removal efficiency and mineralization rate increased from 23.2–87.1% and 5.3–27.9% to 81.7–100% and 15.9–51.3%, respectively, when the barrier thickness reduced from 3 to 1 mm. With the increase of discharge gap, the breakdown voltage, discharge power, gas temperature and residence time increased, while the discharge intensity decreased. The reactor with the smallest discharge gap (3.5 mm) exhibited the highest toluene removal efficiency (78.4–100%), mineralization rate (15.6–40.9%) and energy yield (8.4–18.7 g/kWh). Finally, the toluene degradation pathways were proposed based on the detected organic intermediates. The findings can provide critical guidance for designing and optimizing of DBD reactor structures. [ABSTRACT FROM AUTHOR]

Natural radiation is an integral component and obligatory factor in the self-development of the biosphere. There are two sources of natural radiation on Earth: radioactive elements in minerals of the Earth's crust and inner geospheres and radiation belts, the existence of which was established in the late 1950s. The coevolution of living objects with natural factors, including ionizing radiation, occurs according to the mechanism of homeostatic self-regulation with the participation of negative feedback. Outside the zone of homeostasis, any living species falls into supercritical zones with positive feedbacks, where deviations from a stable state can become irreversible. The effect of ionizing radiation on biological objects has not been sufficiently studied, which is why existing radiation safety standards still do not take into account positive feedback. The author believes that it is necessary to develop a modern unified concept of radiation gene-natural coevolution, in which the radioecogeochemical law of nature, implemented through the periodic-rhythmic evolution of the biosphere, is substantiated from the position of system analysis. In other words, it is proposed to introduce the radiation factor into the prevailing Lamarck–Darwin theory of anthropogenesis as the main driving force. Spatiotemporal coincidence was established between the powerful geodynamic activity of the East African Rift, the operation of a natural nuclear reactor in the area of the Oklo uranium deposit (Gabon), and discharge from our humanoid ancestor—Australopithecus africanus, or afarensis. The article discusses some specific features of human genealogy under conditions of prolonged natural radiation. The dominant role of implicit memory in the awareness of ego stress in the formation of radiophobia is shown. The new direction, undoubtedly, should contribute to solving topical applied issues related to increasing safety in the construction of nuclear industry structures, overcoming radiophobia, increasing the radioresistance of ethnic groups, and optimizing of the regulatory framework of ROSATOM and the IAEA in the field of nuclear energy use. [ABSTRACT FROM AUTHOR]

Abstract—The article considers the efficiency and possibility of introducing into operation of VVER-1000 power units with the lifetime extended owing to the sliding pressure of the second circuit. The history of research by Russian nuclear engineers on VVER lifetime extension is reviewed. It is shown that reduction of fresh steam pressure within the allowed limits will make it possible to release additional reactivity through release of the temperature effect and to generate additional energy during the period of unit operation on the power effect. A mathematical model of changes in the main parameters of the reactor unit, including reactivity balance equations, Stodola–Flugel equation for the turbine, and heat balance and heat transfer equations in the steam generator, is compiled. The expected increase in power generation for modern fuel loadings of VVER-1000 reactors is calculated. The degree of compliance of the proposed mode with the current technological limitations of the large-series VVER-1000 power units is analyzed. The results of the analysis are confirmed by the numerical experiment on the multifunctional analyzer of the VVER-1000 unit modes developed by the National Research Nuclear University MEPhI built on the basis of the ENICAD environment and the certified diffusion neutronic code PROSTOR. Conclusions are drawn about the possibility of introducing the proposed mode into the operational practice of modern VVER-1000 power units. [ABSTRACT FROM AUTHOR]

In this paper, we discuss the issues of handling reactor graphite. The content of 14C in RBMK graphite is compared with the natural content and emissions from other technogenic impacts. The scale of natural fluctuations in the natural content of 14C and their relationship with technogenic impacts are discussed. Possible doses due to the release of 14C from a hypothetical near-surface storage of reactor graphite during normal operation and accidents are estimated. [ABSTRACT FROM AUTHOR]

A liquid fuel reactor based on lithium and beryllium fluoride salts produces a significant amount of tritium as a result of nuclear reactions of the main constituents of the fuel salt with neutrons. At high operating temperatures (600–750°C), tritium can diffuse through the metal walls of the reactor and pipelines, thus affecting the environmental situation. In this article, a method for calculating tritium distribution in this type of reactor is developed and verified. The calculation method is based on differential equations of tritium mass balance in the system elements taking into account the change in the ratio of chemical forms of tritium due to the redox potential of the salt. The calculation method is verified on the experimental data obtained in the 10 MW(t) Li,Be,Zr,U/F Molten Salt Reactor Experiment (MSRE) carried out at the Oak Ridge National Laboratory (ORNL, United States). [ABSTRACT FROM AUTHOR]

In the course of severe accidents (SA) with the core melting, such a phenomenon as the energetic interaction of the melt with the coolant (steam explosion) may occur, which threatens the reactor vessel integrity and can be a trigger for hydrogen explosion. One of the main parameters determining the power of a steam explosion is the conversion ratio CR (another common designation is δ). The CR is the melt energy fraction converted into mechanical work as a result of the postulated steam explosion. In this study, the results of more than 190 experiments simulating the melt–coolant interaction are considered using the prototype corium melt (TROI, FARO, KROTOS-KFC, ZREX, ANL) and the non-prototypical melts (KROTOS Huhtiniemi, MISTEE, SUW, WUMT, MIXA, EXPO-FITS, FITS, ALPHA, WFCI, etc.). On the basis of the results of this study, the dependences of the maximum value of the conversion ratio on the following parameters were obtained: the melt mass reduced to the coolant mass (Mm/Mf); the ratio of the reduced value of the coolant subcooling to the saturation temperature to the reduced value of the melt overheating above the melting temperature to the melting temperature (); the ratio of components in the corium (Zr, UO2, and others). The dependences obtained are supposed to be used to specify existing semiempirical procedures for assessment of the power of steam explosions in order to decrease their conservativeness in the framework of safety justification for operating NPP and those to be designed with VVER reactor plants. [ABSTRACT FROM AUTHOR]

The paper considers the methodological aspects of experiments planning on the irradiation of experimental fuel rods of a VVER-SKD reactor and analyzes the features of the concept of sequential irradiation in fast and thermal reactors. It is shown that the 149Sm accumulated at the irradiation in the BOR-60 plays the role of a burnable absorber at the initial stage of irradiation in the reflector cell of the IR-8 reactor. This absorber quickly burns out owing to large absorption cross section and its concentration gradually comes to equilibrium. Burnup of 149Sm leads to a gradual increase in the linear load of the irradiated experimental fuel rod, followed by a decrease and reaching the equilibrium irradiation regime. This effect must be considered when planning the irradiation regime for experimental fuel elements of the VVER-SKD reactor. [ABSTRACT FROM AUTHOR]

The report presents part of large-scale work on validation of the KORSAR/GP code for the calculation of the parameters of a VVER reactor in power maneuvering regimes. The relevance of the work is due to the novel requirements for NPP power units related to maintaining a flexible schedule of electric load and the feasibility of its operational change, which is required by both Russian and foreign users of VVER technology. Over the past two years, a significant amount of experimental results obtained in testing the daily schedule of maintaining electric load by a high-power VVER reactor unit was accumulated. Ten daily cycles were tested in succession at the beginning and end of the fuel cycle. An important distinguishing feature was the application of an extended range of the secondary-circuit pressure control, which made it possible to use the coolant temperature variation as one of the main methods to affect reactivity. The first stage of the validation work, which was carried out for pretest calculations, enabled checking the qualitative coincidence of the main parameters of a reactor unit with experimental data. This study is devoted to the validation of reactor unit power and the temperature of coolant in the circulation circuits and determination of the quantitative assessment of error. Results of validation are provided for various schedules of power variation of a reactor unit and, in addition, justified conclusions are made regarding the accuracy of the computational model used in the KORSAR/GP code. [ABSTRACT FROM AUTHOR]

The KIR2 software complex for simulating neutronic stationary and nonstationary processes in reactors based on the Monte Carlo method [1, 2] is discussed. The paper presents a brief description of the software complex being developed, its key capabilities, and the features of the architecture. [ABSTRACT FROM AUTHOR]

The uncertainty of quantification of nuclide density is estimated in the calculations of the burnup of UO2 and MOX fuel in the PWR cell due to the error in the neutron flux and uncertainties in the cross sections from the JEFF-3.3 library. It is shown that there is a difference between the constant components of the error for the indicated types of fuel. This difference decreases with increasing fuel burnup depth. [ABSTRACT FROM AUTHOR]