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This article examines the nuclear dimension of the war in Ukraine. It makes two arguments. First, our knowledge of the nuclear world is fundamentally limited and so any inferences we draw from the war in Ukraine should be tentative. Second, what we are observing in Ukraine appears at least consistent with basic and well-established theoretical insights about nuclear weapons: a) that nuclear weapons deter; b) that nuclear weapons enable; and c) that these political effects of nuclear weapons emerge from genuine nuclear risks. This article lays out these insights and demonstrates their consistency with existing empirical evidence from the war in Ukraine. Overall, therefore, it does not appear that the war in Ukraine should yet cause us to fundamentally rethink our understanding of nuclear weapons in major ways.

In the Nuclear Non-Proliferation Treaty’s (NPT) context, irreversibility has been accepted as a critical principle of multilateral nuclear disarmament. Debates within the expert community and during the NPT Review Process, however, recognise the need to better understand the concept and its role in achieving and sustaining a world free of nuclear weapons. Therefore, the principle of irreversibility is still unclear, and the scope of its content is subject to competing interpretations. This paper deals with the concept of irreversibility in the context of nuclear disarmament from an interdisciplinary approach considering its technical, legal, and political aspects. It explicitly explores Latin American views of such a universal concept. The research relies on analyses of relevant information from Latin American experts and compares it with other sources. Focusing on a regional approach and its specificity could help identify ways that local perceptions can effectively feed into the global debate.

This article presents the discussion and conclusions from a workshop hosted by the Verification Research, Training and Information Centre (VERTIC) in February 2023. The workshop convened experts in IAEA safeguards and in nuclear disarmament to discuss the feasibility of applying lessons and tools from the current safeguards system to a future nuclear disarmament scenario. The aim was to probe certain aspects of safeguards which may be relevant in the context of “irreversibility” and evaluate these based on the participants’ knowledge and expertise. The discussion revolved around the following themes: the scope and purpose of verification; the starting point of verification; current verification processes and frameworks such as the State-Level Concept (SLC) and the 1997 Additional Protocol; and the consequences or lack thereof of non-compliance with a future disarmament treaty.

The safeguards system of the International Atomic Energy Agency (IAEA) might inform a future verification regime for global nuclear disarmament that emphasizes irreversibility. This paper explores aspects of a verification mechanism for nuclear disarmament where the IAEA’s experience may be relevant: scope and purpose of verification; starting point and termination of safeguards; verification processes; conclusion-drawing; and compliance.

The paper deals with the influence of the strategy for using separated plutonium from spent fuel of thermal and fast reactors, as well as homogeneous burning of americium in the BN reactor core on the balance of americium in the Russian nuclear fuel cycle throughout the 21st century. The assessment is carried out with the use of mathematical modeling of nuclear materials movement including nuclide transformations throughout the nuclear energy system based on the CYCLE code. Scenario modeling of the americium and Pu-241 accumulation was carried out in Russia’s two-component nuclear energy system model with thermal (VVER) and fast (BN) reactors. In so doing, spent nuclear fuel (SNF) reprocessing was simulated in 2 options: as priority reprocessing of SNF of VVER reactors (1) or SNF of BN reactors (2). In addition to americium accumulation in the system without burning, the accumulation of this actinide was studied taking into account its homogeneous burning in the MOX-fuel of fast reactors at the level of its equilibrium content of ~1%. It has been shown that the priority reprocessing of VVER spent fuel makes it possible to reduce the americium accumulation by the end of the century by ~8 tons, and the effect is achieved by using freshly separated plutonium with short-term cooling, thus, by priority the americium source is eliminated, without directly handling it. Homogeneous addition of americium to the fuel of fast reactors of the BN-1200 type at a level of ~1% makes it possible to stop accumulation of americium in a two-component energy system by 2070, stabilizing it at a level of ~40 tons in the scenario with priority reprocessing of VVER SNF and ~50 tons in the scenario with priority reprocessing of BN SNF.

The VVER-1200/AES-2006 is recognized as a leading Gen III+ nuclear reactor design, meeting stringent international safety standards. This study evaluates the use of a novel PuO2-ThO2 duplex fuel, incorporating weapon-grade plutonium (WgPu) and thorium, for a hypothetical VVER-1200 assembly. The research also explores incorporating minor actinides (MAs) for transmutation, comparing two methods: MAs coated on WgPuO2 and MAs mixed with WgPuO2 as integral fuel burnable absorber rods. Neutronic properties of these fuels are compared to those of LEU-fueled assemblies. The results show a 135% higher burnup for the duplex fuel compared to LEU, with extended criticality, reduced reactivity swings, and lower Pu-239 concentrations upon discharge. While Np-237 and Am-241 concentrations decrease, Am-243, Cm-244, and Cm-245 increase, but overall radiotoxic waste is reduced. Enhanced safety coefficients are also observed, within acceptable LWR ranges.

With its passive characteristics, anti-single point failure characteristics, and simple system, the heat pipe reactor has become a hot research topic in advanced nuclear power. The loading criticality is an essential and important link before the heat pipe reactor is put into operation. Different from the traditional modular design of heat pipe fuel element, the new heat pipe reactor uses bulk fuel, and the traditional extrapolation fuel element/assembly and other extrapolation criticality experimental techniques have poor adaptability. Different from the KRUSTY reactor, the new heat pipe reactor has four side reflectors, and simultaneously lifting four side reflectors does not meet the regulations, which makes the new heat pipe reactor face safety challenges in the process of loading and extrapolation criticality. Therefore, combined with the characteristics of the reactor, the scheme of assembling bulk fuel was firstly determined, and gradually lifting the side reflectors to reach the criticality. Because of the large reactivity of bulk fuel, the core assembly process may face critical safety risks. To solve this problem, the critical safety in the core assembly process was calculated and analyzed by using Monte Carlo calculation program, and the safe and controllable scheme was determined. In the extrapolation process, the neutron fluence rate calculated by the detector is very important, and the size of the new heat pipe reactor is small, and the space available for the arrangement of the neutron source and the detector is limited. At the same time, the new heat pipe reactor is a fast reactor, and the neutron calculated by the detector is a thermal neutron. In order to solve the statistical influence of the neutron source and the detector on the neutron fluence rate in the extrapolation process, the Monte Carlo calculation program was used to determine the arrangement of the neutron source and the detector as well as the thickness of the detector cladding. In order to meet the principle of half addition, the new heat pipe reactor was proposed to use the extrapolation of the side reflector volume to reach the critical point. However, the direct extrapolation of the side reflector volume caused the convex problem on the extrapolation curve. The extrapolation criticality experimental technology was studied, and a modified method was proposed to weight the volume by taking the reactivity value of the side reflector as the weight factor. The extrapolation process was simulated by using this technology, and the feasibility was verified, which provides an important guidance for the loading and extrapolation criticality of the new heat pipe reactor.

At present, ice plug technology is often used in the maintenance process of nuclear power plants to isolate part of the main pipeline and use refrigerant such as liquid nitrogen for ultra-low temperature treatment, so that the internal fluid is frozen to form ice plug to complete the maintenance of low water level valves. Generally, the design of pipeline materials for nuclear power plants is mainly aimed at the working conditions of high temperature nuclear power plants, and the design of ultra-low temperature environment of liquid nitrogen is less considered, which is easy to cause deterioration of pipeline performance and even lead to broken accidents. Therefore, it is necessary to study the ultra-low temperature thermal fatigue performance of pipelines. In order to investigate the influence of freeze-thaw-heating process on Z2CND18.12 austenitic stainless steel for primary circuit pipes during ice blockage operations, 20 cycles of ultra-low temperature thermal fatigue （−196 ℃→0 ℃→350 ℃→room temperature）simulation experiments were conducted. In this paper, Z2CND18.12 nitrogen-controlled austenitic stainless steel used in a nuclear power plant was selected. In the test, boric acid water was injected into the pipeline and frozen at liquid nitrogen (−196 ℃) to make the internal fluid form an ice plug, and the ice plug was maintained for 96 h. Then it was naturally thawed to room temperature, heated after draining the boric acid water, kept at 350 ℃ for 12 h, and cooled to room temperature again with the furnace. So far, a weekly ultra-low temperature thermal fatigue experiment was completed. The ultra-low temperature thermal fatigue test samples with cycles of 1, 3, 5, 10 and 20 were obtained by repeated hot and cold cycles. After the test was completed, the changes in macroscopic mechanical properties of the materials were evaluated by means of room temperature stretching, 350 ℃ stretching, room temperature impact and microhardness test. Microstructure observation techniques such as metallography, SEM, TEM and XRD were used to study the microstructure of the materials after ultra-low temperature thermal fatigue treatment. The result shows that after 20 cycles of ultra-low temperature thermal fatigue, the Z2CND18.12 austenitic stainless steel pipe still maintains the strength, plasticity, and toughness of the original sample. The microstructure dose not undergo phase transformation or precipitation, but the dislocation density increases and the grain size slightly decreases. It indicates that the microstructure of Z2CND18.12 austenitic stainless steel pipe is stable, with excellent mechanical properties, and can resist the repeated effects of deep cold and hot cycles.

The development of thorium-based nuclear fuel is of significance for the long-term sustainable development of nuclear power. Among these, Th1−xPuxO2 stands as a highly promising nuclear fuel with proven applications in various reactor types. This study aims to explore the structural and qualitative characteristics of thorium-based mixed oxide (Th-MOX) fuel Th1−xPuxO2. With the substitution of Ce for Pu, the sol-gel method was employed to synthesize Th1−xCexO2 (x=0, 0.25, 0.50, 0.75, 1) at varying sintering temperatures (800, 1 000, 1 200, 1 400, 1 600 ℃). The morphology and structure of Th1−xCexO2 were characterized by X-ray diffractometer (XRD), scanning electron microscope (SEM) and Raman spectrometer. The lattice constant and electronic density of state (DOS) of Th1−xCexO2 were calculated using the first-principles calculation software VASP based on density functional theory. The experimental findings demonstrate that the prepared Th1−xCexO2 material exhibits excellent properties in terms of density and particle uniformity. With the increase in sintering temperature, the results of density testing indicate that the material’s density exhibits a trend of initially increasing and then decreasing, ultimately reaching a maximum of 98.4% of the theoretical density. This phenomenon occurs due to as the sintering temperature increases, the diffusion process induces the migration of grain boundaries, enlarging the contact area between grains and reducing the presence of pores, thereby increasing the material density. However, further elevating the sintering temperature results in the formation of a liquid phase on the material surface, impeding the migration of grain boundaries. This leads to an expansion of gaps between particles, an increase in the number of pores, and ultimately a reduction in density. The XRD analysis reveals that variations in sintering temperature have a negligible effect on the lattice constant of Th1−xCexO2. Nevertheless, the grain size demonstrates a trend of initial enlargement followed by reduction as the sintering temperature increases, indicative of a migration-driven process. SEM images reveal a uniformly dense structure in samples sintered at 1 400 ℃. In comparison to CeO2 and ThO2, the Raman spectrum of Th1−xCexO2 (x=0.25, 0.50, 0.75) reveals an additional, relatively weaker peak at around 582 cm−1, suggesting the occurrence of oxygen vacancies within the lattice during the formation process of the mixed oxide. In addition, as the Th content increases in the mixed oxide, there is a reduction in grain size, an expansion of the bandgap, and a narrowing of the conduction band width, the differences in atomic radii contribute to an increase in the average bond length of the lattice constant, Th—O bond and Ce—O bond. The lattice constant of Th1−xCexO2, as determined through VASP calculations, closely match experimental value and follow Vegard’s law. The local DOS (LDOS) of Th and Ce atoms increases with the increase of the element content. Near 2.2 eV, the LDOS of Th atom is very low but Ce atom is very large due to the contribution of Ce 4f state electrons. This study offers a practical approach for the preparation of Th-MOX fuels.

Neutron imaging can nondestructively detect the composition and structure of a sample, and has been used in many fields for quantitative analysis. Neutron imaging can be used to detect elements that are difficult to detect with X-rays, such as hydrogen. When neutron imaging is used for quantitative analysis of hydrogen-containing samples, the neutron beam decays because of scattering with hydrogen rather absorption. However, neutrons scattered in the direction of the neutron beam may still reach the detector, and these scattered neutrons can bring additional intensity to the sample and its edges, bring artifacts to the image, and cause serious interference to quantitative analysis. The elimination of scattering effect is one of the most important problems in the quantitative analysis of neutron imaging. Because samples with hydrogen have a larger total scattering than absorption cross section, the influence of the scattered neutrons on the transmitted image is not negligible. The results of neutron imaging quantitative analysis are also affected by the energy spectrum effect. Since the incident neutron beam is not unienergetic, the probability of interaction between the lower energy neutrons and the sample is high, resulting in the macroscopic cross section of the neutron varying with the thickness of the sample. Several existing scattering correction methods, such as increasing the distance between the sample and the detector, the point scattered functions method, using microchannel plates, and using the blackbody grids, are not applicable to all samples, or not applicable to dynamic neutron imaging, or will increase exposure time or complicate data processing. A new scattering correction method, the double-detector method was presented. The double-detector method used two detectors at two locations to image. The close detector can obtain a high-resolution image, but the image is affected by scattered neutrons. The far detector obtain images that are unaffected by scattered neutrons, but with poor resolution. The scattering neutron distribution can be obtained by processing the two images. By subtracting the scattering neutron distribution from the close image, an image with high resolution and free from scattering neutrons can be obtained to complete scattering correction. The method was used to correct the water steps sample images. And the macroscopic cross section was corrected by the energy spectrum correction factor. The results show that the thickness of the water steps in the close-position image differs greatly from the actual thickness due to the influence of scattered neutrons. And the thickness of the water steps corrected by the double-detector method is in good agreement with the actual thickness. The double-detector method is a feasible scattering correction method.

Neutron imaging is a transmission imaging method, which has a wide range of applications in the fields of non-destructive testing of internal defects in materials, the distribution of light elements in samples, and the detection of radioactive samples due to the advantages of neutrons with deep penetration, ability to distinguish isotopes, and sensitivity to light elements. With the continuous development of various application fields, the requirements for the spatial resolution of neutron imaging are getting higher and higher, and improving the spatial resolution of neutron imaging and continuously approaching the micron level has become a key research direction in the field of neutron imaging. At present, the international advanced neutron imaging device has been able to complete the high-resolution neutron imaging test with spatial resolution at the micron level, and the domestic research in this field has also made some progress with the completion of the three major neutron sources, and is gradually approaching the international advanced level. The thermal neutron imaging device and the cold neutron imaging device of the China Advanced Research Reactor have been completed and put into use in recent years, and the maximum neutron flux density at the sample of the device can reach 4×109 cm−2·s−1, the optical path system has been greatly improved compared with the original neutron imaging device, and the spatial resolution of the detector has reached the sub-millimeter level, which has been widely used in aerospace, cultural relics and archaeology, hydrogen content characterization and other fields. The main factors affecting the spatial resolution of neutron imaging in neutron imaging devices are neutron sources, neutron optical path systems and neutron imaging detectors. In this study, based on the relevant test requirements in the fields of zirconium alloy hydrogen content and cultural relics production technology, the influencing factors of neutron imaging spatial resolution were studied, and a set of high-spatially resolved neutron imaging detectors based on the thermal neutron imaging device of China Advanced Research Reactor was built. At the same time, the spatial resolution of the detector was calculated theoretically, and the detector was debugged as a whole. Finally, the relevant band bundle commissioning of the detector system was completed on the thermal neutron imaging facility of China Advanced Research Reactor, and the syringe needle samples were tested. The test results show that the spatial resolution of the detector reaches 11.3 μm, and the actual available field of view size reaches 23.6 mm×28 mm, achieving a spatial resolution of 10 μm.

As radioactive sources are increasingly used in scientific research, industry and other fields, radioactive sources are often lost due to improper management and control. Since gamma rays have higher penetrating power and longer action distance than α particles and β rays, they can affect a larger area and have radiation effects on various substances in the environment. Therefore, the search and positioning of radioactive sources has always been the focus of radioactive source search. Although gamma detection equipment can achieve relatively precise positioning of a single radioactive source, when there are multiple radioactive sources in the radiation environment, radioactive sources with weaker energy can easily be ignored. In order to enable the gamma camera to accurately locate multiple radioactive sources in the unknown radiation field at the same time, this paper proposed a multi-distance system matrix superposition gamma image reconstruction algorithm based on the maximum likelihood expectation maximization (MLEM) algorithm. First, Monte Carlo simulation software was used to build a gamma camera simulation, and the impact of different encoding patterns on detection efficiency and the impact of radioactive sources at different distances on detection performance were analyzed. In order to be as close to the actual situation as possible, the photoelectric effect, Compton scattering, electron pair effect and decay of the radioactive source were considered in subsequent simulation experiments, and the simulation structure is consistent with the gamma camera structure. Then, simulation was used to obtain projection data, and different distance system matrices were calculated based on the impact of radioactive source distance on detector collection efficiency. Besides, in order to simultaneously located multiple radioactive sources within different distance ranges, a new system matrix was formed by superimposing multiple system matrices with different distances, and then calculated with the projection data, and through continuous iteration, a higher quality reconstructed image and the position information of multiple radioactive sources were obtained. Finally, in order to verify the feasibility of the algorithm, the algorithm was used for actual verification of 60Co and 137Cs radioactive sources in a real environment, accurately reconstructed the position information of the two radioactive sources. Experimental results show that this method can quickly locate radioactive sources. It has high convergence when using a single source, and it can accurately locate multiple radioactive sources when using multiple sources. It has a longer imaging distance and higher positioning accuracy than related decoding algorithms. This method is practical and feasible in practical applications.

Cathodes serve as sources of electron beams, and their performance plays a great part of beam quality in electron accelerators. The introduction of surface plasmon polaritons (SPPs) into different photocathodes has been investigated to increase their quantum efficiency. Currently, the SPPs excitation is predominantly achieved in the infrared and visible bands due to limitations in the permittivity of common noble metals. Exciting SPPs in the ultraviolet band typically involve attaching aluminium nano-particles to the material surface, which can impact the optical properties of the material. Therefore, it is important to develop a method for exciting SPPs in the ultraviolet band without affecting optical properties of the material. The cesium telluride (Cs2Te) photocathode represents a lifetime of several months, a quantum efficiency of up to 10%, a response time of less than 1 ps, a thermal emittance of less than 1 mm·mrad/mm, and moderate operating vacuum requirement of 10−7 Pa. These characteristics make Cs2Te an exceptional photocathode material for generating high-current electron beams. However, a drawback of Cs2Te photocathodes is that they need to be excited by high-power ultraviolet lasers, which are more challenging to obtain compared to green lasers. Consequently, there is a continued demand for increasing the quantum efficiency of Cs2Te to reduce the laser power requirements. In this paper, accuracy of the Monte Carlo model to simulate the Cs2Te photoemission was verified. SPPs were introduced into Cs2Te photocathode by means of micro-nano grating structure, which increases substantially the absorption of incident photons and hence form a local field enhancement to regulate initial photoelectrons distribution. The Monte Carlo model was modified according to the theory of SPPs to simulate the photoelectric emission process of the Cs2Te photocathode after the introduction of SPPs. The simulation results show that the quantum efficiency of Cs2Te photocathode increases by 60% while the thermal emittance is basically unchanged, with the introduction of SPPs. This is due to the photoelectrons population, resulting from the additional laser absorption induced by the SPPs, decreases exponentially from the Cs2Te/Al interface to the vacuum/Cs2Te interface. The majority of these increased photoelectrons are distributed near the deeper interface, resulting in higher scattering probabilities during their transport from inside the photocathode to the surface, leading to a reduction of thermal emittance of the emitted electrons. Combined with conventional methods for controlling the thermal emittance of photocathodes, such as reducing the energy of incident laser photons and operating at low temperatures, it is anticipated that a photocathode with higher performance, characterized by low thermal emittance and high quantum efficiency, can be achieved. This advancement holds significant implications for the future development of high-brightness electron sources.

To improve the analytical performance of low resolution gamma spectrometers, a new method based on spectrum deconvolution to detect weak peaks was put forward. First of all, with SuperMC Monte Carlo simulation code, a geometric model matching the measured conditions was established, and then the system response matrices of a NaI(Tl) spectrometer was constructed. Secondly, in order to ensure the accuracy of the gamma spectrum deconvolution, the nonlinear least square spectrum stabilization technique was applied to stabilize the measured spectra. Thirdly, the long time measured background spectrum was used for background deduction. And then the L-R deconvolution algorithm based on the system response matrices was used to improve the saliency of weak peaks. Finally, the symmetric zero-area transformation method was used to detect the weak peaks. To verify the feasibility and applicability of the method, the gamma spectra of weak source 137Cs disturbed by the Compton scattering continuum of high activity 60Co were simulated, and the gamma spectra of weak 137Cs and 152Eu sources disturbed by the Compton scattering continuum of high activity 60Co were measured, as well as the gamma spectra of weak source 137Cs. The simulated and measured results show that the significance of the weak peaks of 137Cs and 152Eu is markedly improved after deconvolution, and the peak-finding success rate increases greatly. Specifically, the peak-finding success rate of the weak peak 137Cs increases from 25.9% to 42.0%, 73.4% to 90.8%, and 61.4 % to 75.2% when it is overwhelmed by the environmental background, the Compton scattering of 60Co and the statistical fluctuation, respectively. The peak-finding success rate of all weak peaks of 152Eu disturbed by the Compton scattering continuum of different degrees, are improved to varying degrees, with an increase of 6.0% to 97.0%. Especially, the peak-finding success rate of 295.9 keV weak peak of 152Eu is significantly elevated from 3.0% to 100%, as well as 1 528.1 keV weak peak of 152Eu from 6.1% to 97.0%. Therefore, it is proved that the method based on spectrum deconvolution to detect weak peaks can effectively improve the weak peaks detection capability of low resolution spectrometers. The performance of the method is restricted by the error of the system response matrices, which is attributed to the mismatch between the Monte Carlo geometric model and the measured conditions, so the method is mainly suitable for the fixed measurement in the laboratory. Furthermore, this method is not only applicable to NaI(Tl) measurement system, but also to other spectrometers such as CsI(Tl), LaBr3(Ce), CdZnTe and HPGe, and is also suitable for application scenarios with shielding measures and anti-coincidence techniques. But, the system response matrices matching the measurement system needs to be established and applied.

To establish the national primary standard for beta radiation tissue absorbed dose and to achieve international equivalence, the National Institute of Metrology (NIM) of China participated in the BIPM KCDB: EURAMET.RI(I)-S16 comparison that organized by EURAMET (European Association of National Metrology Institutes) and coordinated by the Physikalisch-Technische Bundesanstalt (PTB) in 2019. The circulation of the chamber and of the electronic measurement system will be effected in a star pattern, with both devices being sent back by every five participants to the pilot laboratory for testing. 16 institutes from 16 different countries will calibrate the comparison device according to their quality system in beta reference fields of the radionuclides 147Pm, 85Kr, 90Sr/90Y and 106Ru/106Rh. NIM calibrated the transfer ionization chambers using its beta radiation tissue absorbed dose primary standard, determining the calibration factors for Hp(0.07) at incident angles of 0°, 45°, and 60° in the radiation field of 85Kr and 90Sr/90Y sources, as well as the calibration factors for Hp(3) at various angles for the 90Sr/90Y source. The calibration factors obtained in this work were compared with the reference values provided by EURAMET. The relative deviations of the calibration factors for Hp(0.07) at incident angles of 0°, 45°, and 60° from the 90Sr/90Y and 85Kr sources compared to the reference values are better than ±0.8%; The relative deviations of the calibration factors for Hp(3) at incident angles of 0°, 45°, and 60° from the 90Sr/90Y source compared to the reference values are better than ±0.9%. The relative deviations for the nine results are better than ±1%. These results were incorporated into the International Bureau of Weights and Measures (BIPM) Key Comparison Database (KCDB), achieving international mutual recognition for the personal dose equivalents Hp(0.07) and Hp(3) in beta radiation.

14C, a radionuclide, which is widely used in all works of life, such as medical research and determination of PM2.5, has been realized autonomous production in China. It relied on imports, which was expensive and unstable supply in the past time. At present, Qinshan Nuclear Power has started 14C production to completely solve the problem of the serious shortage of 14C independent supply in China. Aluminum nitride powder is the main raw material for the preparation of 14C, whose elemental aluminum directly affects the safety and quality of 14C isotopic targets in nuclear reactors. Aluminum nitride is very stable in an inert high temperature environment. But a lot of elemental aluminum will react with water producing hydrogen in a high temperature environment which can result in the expansion of component rods or it will react with cladding tubes producing hydrogen embrittling effect, and these can affect the quality of 14C isotopic target. According to the characteristic of aluminum nitride insoluble in acid, this study researched different dissolution effect of elemental aluminum in aluminum nitride nano-powder by nitric acid, hydrochloric acid and sulfuric acid. It was found that the above mentioned acid can only dissolve a small amount of aluminum nitride powder, and there was a small amount of aluminum nitride powder suspending in solution, which would seriously interfere the determination of elemental aluminum. At present, microwave digestion, laser ablation, X-ray fluorescence and other methods are used to dissolve or determinate aluminum nitride, but they are not suitable for the determination of elemental aluminum. The determination of elemental aluminum is generally used in pharmaceutical, soil, tea and other samples, but most of the aluminum in the above samples exists as an impurity element, or aluminum is combined with organic matter, which is easy to separate from elemental aluminum. However, it is difficult to achieve complete separation of elemental aluminum(unbound elemental aluminum) and the aluminum nitride matrix in the aluminum nitride powder. This study used replacement reaction to selectively dissolve elemental aluminum. Then the suspended aluminum nitride powder was separated by filter membrane filter. By this method, it could solve the determination interference to elemental aluminum. This study established the determination of elemental aluminum in aluminum nitride nanoparticles by inductively coupled plasma mass spectrometer(ICP-MS), which has the advantages of low determination limit, high accuracy, fast determination speed and small matrix effect. This method’s linear range is 10-100 μg/g and the relative standard deviation is 3.8%(n=6). Using high purity alumina powder to verify the established determination method, the dissolution efficiency can reach 96%-98%.

Brachytherapy can kill cancer cells more effectively and make the normal tissues around the tumor free or less irradiated. It has become an effective means for the treatment of malignant tumors. 103Pd has a half-life of 16.96 days and a significant advantage in the field of brachytherapy due to its unique decay properties. In order to achieve the large-scale production of 103Pd, the preparation process based on Cyclone-30(C30) cyclotron was studied. The rhodium metal was electroplated on the copper target base by pulse electroplating method. The rhodium-plated target was transferred to C30 solid target station for irradiation with beam energy of 16-18 MeV, and the beam current was 200 μA. In order to obtain Curie-level 103Pd, the beam integral should be more than 10000 μA•h. After irradiation, the target was transferred to the separation and purification hot cell, the rhodium plating was separated from the target base first, then was ground into powder, and dissolved by high temperature melting of potassium bisulfate. The rhodium was converted into soluble rhodium sulfate, and the 103Pd nuclide was separated and purified by AG1-X8 resin. The impurities of Rh, Fe, Cu and Zn were rinsed with 6 mol/L hydrochloric acid and 0.03 mol/L hydrochloric acid, respectively. At last, 103Pd was desorbed with a mixed solution of ammonium chloride-ammonia mixed solution(volume ratio 1∶1). The radioactive activity, radionuclide purity and specific activity of 103Pd were measured respectively. The preparation of sealed seed core was studied by using the prepared 103Pd feed solution. The results show that the rhodium plating layer is smooth and dense, and firmly bonded with the copper plate, the rhodium layers mass thickness is greater than 150 mg/cm2. The production capacity of 103Pd is greater than 37 GBq, the radionuclide purity is greater than 99.9%, and the specific activity is greater than 875 GBq/mg, and activity concentrations greater than 6.2 GBq/mL. The preparation field of 103Pd brachytherapy sources is more than 90%, it can meet the requirements. The 103Pd preparation process is stable, and the quality is controllable. The large-scale production capacity has been reached, and it provides a stable source for the research of 103Pd brachytherapy sources.

Inert gas argon has important applications in the fields of daily life, industrial manufacture and nuclear environmental monitoring. The key to its application is to selectively and efficiently separate and purify Ar from complex gas components. However, the separation of O2 and Ar at normal temperature is full of challenge due to their similar molecular sizes and adsorption properties as well as their close boiling point. In industry, high purity Ar is mainly produced by low temperature rectification, which is mature technology, but the equipment is large, high-energy consumption and explosive. Adsorption separation is a common method of gas separation. Molecular sieves are the most widely used material due to their huge specific surface area, aboudant pore structure and excellent adsorption properties. The silica tetrahedron in molecular sieve is electrically neutral, while the aluminum tetrahedron is electronegative, which makes it necessary to use H+ or Na+ outside the molecular sieve skeleton to maintain electrical neutrality. In actual application, the interaction between adsorbent and adsorbent can be enhanced by ion-exchange modification of molecular sieve material, so as to effectively improve the O2/Ar separation capacity of molecular sieve. In this paper, the application of ion-exchange modified molecular sieve materials in the field of O2/Ar adsorption separation was systematically summarized and reviewed. The results show that the O2/Ar separation coefficients of molecular sieves modified by alkaline earth metal ions like Ca2+, Sr2+ and Ba2+ can reach about 2.0, while that of Ce3+-exchanged X-type molecular sieve is as high as 5.9, it is an ideal adsorption material for O2/Ar separation. However, the high separation coefficient is limited to the range of low partial pressure, and the O2/Ar separation coefficient under normal pressure is below 2.0. The Ag+-exchanged molecular sieves, namely silver-loaded molecular sieves, have excellent Ar adsorption selectivity. At present, the O2/Ar separation coefficients of silver-loaded molecular sieve materials at atmospheric pressure are around 2.0, but the preparation cost of silver-loaded molecular sieves is high, and it is more suitable for small-scale separation applications, it means that the efficient separation of O2 and Ar at normal temperature and pressure is still challenging. The development of novel ion-exchange modified molecular sieve materials with high adsorption capacity and high adsorption selectivity for the separation of O2 and Ar is still one of the focuses and trends of future research projects.

In order to accurately simulate the large-scale migration of plutonium in groundwater(a three-phase system), it is necessary to understand the interaction mechanisms between plutonium(Ⅳ), natural colloid, and granite(immobile media). The experimental results show that plutonium pseudo-colloids(i.e., Pu(Ⅳ) sorbed on the surfaces of soil colloids) have the strong mobility(i.e., the recovery percentage of plutonium after plutonium pseudo-colloids transported through columns filled with pore media, R). The mobility R increases with soil colloid concentration ρ(0≤ρ≤375.4 mg/L, 1.3%≤R≤52.5%). The master complexation speciation of Pu(Ⅳ) on the surfaces of soil colloids is \begin{document}$\equiv {\text{SOPu}} $\end{document}(OH)3. The speciation percentage(＞95%) is positively correlated with the colloid concentration. The results of experiments and surface complexation calculations confirm that plutonium pseudo-colloids enhance the mobility of Pu(Ⅳ) and thus pose a greater risk of contaminating far-field waters than the dissolved Pu(Ⅳ). The calculations of DLVO interaction potential indicate that the trough depth of the second energy minima(Фmin2) between colloids and media deepens with increasing colloid concentrations. This attractive interaction potential results in an increase in the adsorptive deposition efficiency of plutonium pseudo-colloids on the media surfaces. The DLVO calculations are consistent with the observation: the mobility of plutonium pseudo-colloids become weak(375.4 mg/L＜ρ≤2017.8 mg/L, 52.5%＜R≤12.7%).

Radioactive TBP/kerosene(OK) organic waste liquid is produced in the process of uranium purification and spent fuel retreatment. In order to avoid the safety risks caused by the accumulation of radioactive organic waste liquid over the years, it is urgent to treat it. In this paper, the conventional Fenton oxidation method and microwave synergistic Fenton method were used to treat simulated radioactive TBP/OK waste liquid, and the effects of mass fraction of Fenton’s reagent ρ(\begin{document}$ {\text{Fe}}^{\text{2+}} $\end{document}) and ρ(\begin{document}$ {\text{H}}_{\text{2}}{\text{O}}_{\text{2}} $\end{document}), initial pH and oxidation time(t) on the removal of chemical oxygen demand(COD) was experimentally investigated. The results of the conventional Fenton oxidation method show that: at the optimal process conditions, ρ(\begin{document}$ {\text{Fe}}^{\text{2+}} $\end{document})=20 mg/L，ρ(\begin{document}$ {\text{Fe}}^{\text{2+}} $\end{document})/ρ(\begin{document}$ {\text{H}}_{\text{2}}{\text{O}}_{\text{2}} $\end{document})=1∶10, initial pH=1.90，t=15 min, the COD removal can reach up to 72.01%; the most important factor is the ratio of ρ(\begin{document}$ {\text{Fe}}^{\text{2+}} $\end{document})/ρ(\begin{document}$ {\text{H}}_{\text{2}}{\text{O}}_{\text{2}} $\end{document}), followed by ρ(\begin{document}$ {\text{Fe}}^{\text{2+}} $\end{document}), initial pH, and oxidation time. The microwave synergistic Fenton method is superior to the conventional heating Fenton method, and even better than the conventional Fenton oxidation method, when under the same optimal reaction conditions. As for the microwave synergistic Fenton method, the COD removal can be increased to 95.80%. The dosage of Fenton reagent is also reduced from ρ(\begin{document}$ {\text{Fe}}^{\text{2+}} $\end{document})=20 mg/L，ρ(\begin{document}$ {\text{H}}_{\text{2}}{\text{O}}_{\text{2}} $\end{document})=200 mg/L to ρ(\begin{document}$ {\text{Fe}}^{\text{2+}} $\end{document})=10 mg/L，ρ(\begin{document}$ {\text{H}}_{\text{2}}{\text{O}}_{\text{2}} $\end{document})=100 mg/L. It can be seen that the microwave synergistic Fenton method not only further improves the COD removal, but also reduces the dosage of Fenton reagent.

In recent years, semiconductor photocatalytic methods have made great progress in the reduction and extraction of uranium from water. In recent years, due to its advantages of less pollution, low energy, and high material reuse efficiency, photocatalysis has become an important direction in the study of reduction and extraction of uranium in water, and it also has the potential to treat nuclear wastewater and extract uranium from seawater. This article reviews the mechanism of photocatalytic reduction of uranium, and the characteristics, synthesis and classification methods of catalyst materials. In order to make the classification more intuitive and specific, classification is based on composition rather than structure, and it is divided into titanium dioxide and its modified composite materials, graphite phase carbon nitride and its derivatives, bismuth photocatalysts, metal sulfide photocatalysts, etc. The advantages and disadvantages of different photocatalysts were analyzed and a series of comparisons were made to provide reference and basis for research on uranium extraction through photocatalytic reduction.

The separation and recovery of uranium from natural water is of great significance to the exploitation of unconventional uranium resources. In our paper, a novel polyacrylonitrile(PAN)/Hummers oxidized lotus seedpod carbon composite was prepared, and adsorption of U(Ⅵ) ion from U(Ⅵ)-CO3 solution under weakly alkaline condition was also studied. Batch adsorption experimental results show that at room temperature, pH=8.0, ρ(U(Ⅵ))=50 mg/L, cT(CO3)=2 mmol/L, the adsorption reaches equilibrium after 840 min, and the maximum adsorption capacity is about 51.92 mg/g. The increase of the pH values, calcium ion concentration, magnesium ion concentration and carbonate concentration inhibit the U(Ⅵ) adsorption. Thermodynamic results indicate that it is a spontaneous, endothermic chemical adsorption process. Characterization results show that the composite displays good structure stability during U(Ⅵ) adsorption process, meanwhile, the phenolic —OH and —COOH(C=O) are the main adsorption sites on the composite surfaces. In the presence of calcium/magnesium ions, the order of adsorption affinity of U(Ⅵ) on the composite material is as follows: Mg-U(Ⅵ)-CO3 system＞Ca-U(Ⅵ)-CO3 system. The results can provide the new materials for the separation and extraction of uranium from natural uranium-containing water.

Uranium-zirconium alloy is an important nuclear fuel in Integral Fast Reactor, which is of great significance to study its basic physical properties at high temperature by using advanced calculation methods. This work used deep potential molecular dynamics method, which combines the high accuracy of the first principles with the high efficiency of the classical molecular dynamics, to perform an evaluation of the static and thermophysical properties of body-centered cubic phase zirconium, uranium, and uranium-zirconium alloy. Firstly, the deep potential(DP) models of body-centered cubic zirconium(Zr-BCC), body-centered cubic uranium(U-BCC), and body-centered cubic uranium-zirconium alloy(U-Zr（BCC）) were trained by using deep neural network machine learning. Secondly, the DP models were used to predict equilibrium state equation, lattice constant, elastic properties, and phonon spectrum of the three systems, and the predicted results can reach the accuracy of the first principles. Then, the variation of heat capacity and density at constant pressure of Zr-BCC, U-BCC, and U-Zr（BCC） with temperature were predicted by using DP models, and the results are in good agreement with the experimental values. The research results show that the machine learning method provides an important path for successfully exploring more complex nuclear fuel properties.

As a reactor which can reusing low-abundance 235U resources, fast reactor construction is an important part of China’s three-step nuclear energy process. Due to high burnup and strong irradiation of the spent fuel oxides of fast reactor, the dry reprocessing method based on molten salt electrolytic refining has unique advantages such as the ability to handle the widest range of fuel types and higher stability of the reaction system. Before the electrolytic refining step, the oxide spent fuel needs to be reduced to metal through electrical or thermal reduction. Thermal reduction requires the use of calcium, lithium, and magnesium as reducing agents. Compared with other alkali metal reduction process, lithium reduction process at high temperature has the advantages of simple recovery process, moderate temperature and lower equipment requirements, it may become the future tendency of thermal reduction. In this paper, the research progress of lithium high-temperature chemical reduction of spent fuel in dry reprocessing and compared the characteristics of lithium thermal reduction processes in various countries was investigated. The mechanism of high-temperature reduction with lithium was deduced based on thermodynamic calculation, and the influencing factors of reaction were discussed from the perspective of kinetics. Finally, the relevant research on the electrolytic recovery of residual salt from thermal reduction was described. The different reaction routes and technological process of lithium thermal reduction were summarized. The shortcomings of existing research were summarized, and the trend of lithium thermal reduction in the future was prospected. It gave a reference for the research of the process flow in China’s dry reprocessing.

Due to anthropogenic and long-term geological factors, radionuclide leakage may occur in deep geological repositories of radioactive waste when groundwater infiltrates a failed waste canister. The released radionuclides can migrate to near and far field regions through groundwater, posing a serious threat to the ecological environment and human health. As a consequence, knowledge of radionuclide speciation in groundwater is essential for understanding their transport behavior. Uranium is one of the key elements of significant concern for deep geological disposal of high-level radioactive waste. In this study, the uranium species in groundwater from boreholes BS03(pH=9.43) and BS05(pH=7.51) located at the Beishan area(Gansu province) were calculated by the hydrogeochemical software PHREEQC(version 3). The effects of pH, temperature, and initial uranium concentrations on ternary uranyl species are presented. The results indicate that UO2CO3(aq) and (UO2)2CO3(OH)\begin{document}$_3^- $\end{document} are the dominant species in a mildly acidic condition, while CaUO2(CO3)\begin{document}$_3^{2-} $\end{document} and Ca2UO2(CO3)3(aq) are the major species in neutral and weakly alkaline conditions. In BS05 groundwater, the concentration of Ca2+(5.11 mmol/L) is about three times higher than that of Mg2+(1.58 mmol/L). However, the concentration of calcium uranyl carbonate is estimated to be about 250 times higher than that of magnesium uranyl carbonate(MgUO2(CO3)\begin{document}$ _3^{2-}$\end{document}). With an increase of temperature, Ca2UO2(CO3)3(aq), CaUO2(CO3)\begin{document}$_3^{2-} $\end{document} and UO2(CO3)\begin{document}$_3^{4-} $\end{document} concentrations progressively diminish. At uranium concentrations between 4×10−4 and 1×10−1 mmol/L (25 ℃), the dominant uranium species, in order of decreasing concentration are: Ca2UO2(CO3)3(aq)(x＞62%)＞CaUO2(CO3)\begin{document}$_3^{2-} $\end{document}(x≈35%)＞MgUO2(CO3)\begin{document}$_3^{2-} $\end{document}/UO2(CO3)\begin{document}$_3^{4-} $\end{document}(x≈1%), irrespective of the concentrations of Ca2+, Mg2+, and Na+. In the near-neutral and slightly alkaline conditions, the results suggest that the formation of neutral Ca2UO2(CO3)3(aq) and anionic CaUO2(CO3)\begin{document}$_3^{2-} $\end{document} in Beishan groundwater reduces the distribution coefficient between uranium and environmental geo-medium, thereby enhancing the mobility and migration potential of uranium. The research findings contribute to understanding the transport and release mechanisms of uranium, as well as interactions between uranium species and engineered and natural barriers.

The transport and transfer of radionuclides in the ecological environment (referred to migration) have long been of a particular concern in the field of radiological environmental protection. This paper systematically reviews the current relevant research statuses on mechanisms, models and parameters of radionuclide migration in the ecological environment, and the methods for assessing doses of humans and non-human species, as well as the non-equilibrium migration models and parameters of radionuclide migration in case of accident at a nuclear facility. From multi-dimensional perspectives, such as computational models, evaluation methods, and parameters, etc., this paper explores the transport and transfer behaviors of radionuclides in the ecological environment. In terms of radionuclide migration, this paper systematically reviews the transport processes and migration models of radionuclides from source term release to atmospheric, terrestrial, surface water (rivers, lakes, etc.), groundwater, and marine environments. It elaborates on the basic theories of radionuclide transport and transfer in the ecological environment, compares the characteristics of different transport modes, and summarizes the transport and transfer behaviors of radionuclides in the ecological environment. In terms of environmental radiation exposure assessment, this paper summarizes both individual-related and source-related evaluation methods, describing the basic methods of environmental radiation exposure assessment. It presents a schematic diagram of the main steps and dose estimation models for assessing the impact of radionuclides released into the environment, and sorts out the methods, evaluation models, and dose assessment software for exposure assessment from the perspectives of both the human and non-human species, domestically and internationally. Moreover, for environmental radiation exposure assessment in case of accidents, this paper takes tritium and 14C as examples to provide an overview of short-term migration, accident exposure assessment, and real-time consequence evaluation and decision support for nuclear accidents. It proposes suggestions for further research in the aspects of radionuclide marine transport, dose assessment models for non-human species, short-term accidental release, and the application of artificial intelligence, to provide scientific basis and technical support for nuclear and radiation emergency response, environmental monitoring, and risk assessment.

On August 24, 2023, the Japanese government officially discharged nuclear wastewater into the ocean, causing widespread international concern. Radionuclides in nuclear wastewater pose a serious threat to the marine environment and to the national health of neighboring countries. This paper aims to accurately characterize the diffusion and transport processes of radionuclides in the ocean, which is the key to monitoring and prevention of radionuclides effects. Based on regional ocean modeling system (ROMS), this paper set up a high-resolution regional ocean model of the North Pacific Ocean. The distribution of radionuclides discharged into the ocean from Fukushima over a period of five years was simulated by using passive tracer experiments. The relative concentration of passive tracers are capable of representing the radionuclides distributions and pathways. The results show that the radionuclides mainly move eastward with the Kuroshio extensions after being discharged into the sea, but some radionuclides still arrive at the Chinese offshore in a southwesterly direction due to southwesterly return currents, mesoscale eddy activities and mode water subduction processes. Among the material exchange channels between China Seas and open ocean, radionuclides mainly enter the East China Sea through east of Taiwan channel, and penetrate into the South China Sea through the Luzon Strait, while the other channels are all export radionuclides from China Seas to the open ocean. Shelf circulations play a vital role in the exchange of materials between sea areas, which transport radionuclides into the Yellow Sea and the Bohai Sea particularly by Taiwan warm current and Yellow Sea warm current. Besides, seasonal variations of ocean circulations can significantly affect the transport of radionuclides. The ocean currents on the shelf can even show reverse directions between summer and winter. A new assessment of the impact time of radionuclides at different concentration thresholds in the four Chinese Seas and key channels was finally given. The East China Sea is the first area affected by radionuclides, and the Yellow Sea and the South China Sea are affected by radionuclides at similar times, while the Bohai Sea was the latest with lowest concentration of radionuclides. Radionuclide impact times are more homogeneous vertically in the shallow channels, while the impact times of deep-water channels are different in vertical due to the complex vertical structure. The model results systematically provide the three-dimensional transport pathways and dynamical mechanisms for radionuclides entering the China Seas from the North Pacific Ocean, and confirm that the high concentration radionuclides can not reach the China Seas during the five years of the simulation.

A large amount of artificial radionuclides have been released into the ocean, contributing to serious nuclear pollution in marine environment, and arising public concerns and worry around the world. The Fukushima-derived artificial radionuclides can also be used as tracers to reveal the migration, transformation processes, and fate of artificial radionuclides in the ocean. The most polluted port within less than 1 km from the Fukushima Daiichi Nuclear Power Plant (FDNPP) was focused on in this study. The most polluted port near the FDNPP serves as windows to reflect progresses and effectiveness of decommissioning of the FDNPP, which is inaccessible for public and many other counties around the world. Historical activities of 134,137Cs in seawater, marine sediment, and marine fish were reconstructed from April 2011 to October 2023 on the basis of over 1 000 reports from Ministry of Economy, Trade and Industry of Japan, Nuclear Regulation Authority of Japan, and Tokyo Electric Power Company. The patterns of the three-stage evolution of 134,137Cs in seawater, the four-stage evolution of 134,137Cs in sediments, and the three-stage evolution of 134,137Cs in marine fish were proposed to quantify the activity levels and effective half-lives (EHL) of 134,137Cs at different stages. The evolutions of historical 134,137Cs in seawater, sediment, and marine fish were closely related to multiple countermeasures of decommissioning at the FDNPP, including the relocation of the drainage channels during June 2014 to April 2015, seabed covering of port in April 2015, removal of highly contaminated retained water in December 2015, filling of tunnels and towers in December 2015, and completed construction of sea-side impermeable walls in February 2016. The longest EHL of 134,137Cs in marine sediment indicates the memory effect of marine sediment and its persistent and dominated contribution to 134,137Cs in marine fish. Additionally, a highly consistent activity ratio of 134Cs to 137Cs (about 1.0) was simultaneously calculated in seawater, sediment, and marine fish, indicating the transferring of the Fukushima-derived 134,137Cs in multiple matrices in the marine environment. The temporal variation of concentration factor of 137Cs in marine fish was also constructed to reveal the dynamic processes of the enrichment and uptake of 137Cs in marine fish from seawater. The relatively high value of concentration factor of 137Cs in marine fish was observed during the initial period of nuclear accident followed by a decline in concentration factor of 137Cs to about 100 L/kg. This study would provide scientific evaluations for the effectiveness of the decommissioning of the FDNPP and the consequences of Fukushima contaminated water discharged into the ocean.

The migration and fate of pollutants discharged by human into marine environment are mainly controlled by the transport processes of seawater and particulate matters. By utilizing natural or artificial radioactive isotopes, which are widely present in the ocean, the timescales of seawater and particle dynamics can be quantitatively evaluated. During the period of October 2019 to January 2020, the activities of dissolved radium isotopes (224Ra, 223Ra, and 228Ra) in the water body and coastal groundwater in Guanghai Bay, western Guangdong Province, China were investigated in this study. Meanwhile, the activities of particulate nuclides (7Be, 210Pb, and 234Th) of surface sediments and suspended sediments were also investigated in the same sampling period. Based on the collected data of radioactive isotopes, a comprehensive study on the timescales of transport processes of seawater and particulates in Guanghai Bay was carried out. The residence time of water in the bay was calculated by using the improved 224Ra/228Ra model, and the results show that the timescale for the migration and spread of water and its dissolved pollutants carrying to the outer of the bay is around 1 day. Additionally, two box models for the particulate 7Be and 210Pb were constructed, respectively, based on a quantitative assessment of all sources and sinks of both isotopes in the water body. The results reveal that the atmospheric deposition compared to the local river input is one of the most significant sources of 7Be or 210Pb in the study area and the net tidal export from the bay to the open sea constitutes the primary sink for the particulate 7Be in the water body. For the particulate 210Pb, the sediment burial is the major sink when the sedimentation rate is as low as that of islands located outside the bay (i.e., 0.78 cm/a). Whereas the input of suspended sediments from the Pearl River, which is carried by coastal currents of western Guangdong, becomes the most important source when the sedimentation rate increases to 3.0 cm/a of Pearl River estuary. The residence time of particulate matters in the bay was estimated by using the particulate 7Be, 210Pb and 234Th residence time models for typical bays, respectively. The results show that the residence time of particles calculated by 7Be and 234Th was quite similar, approximately 2 days. In contrast, the particulate residence time estimated by 210Pb was much longer, ranging from 12 to 45 days, with shorter particulate residence time corresponding to higher sedimentation rate. Comparing the estimates of residence times for particulate matters and considering the contribution of sediments from Pearl River, the study concludes that the particulate residence time calculated by 210Pb provides a better indication of the timescale for the transport processes of particles and their associated pollutants in Guanghai Bay. The findings of this study are intended to provide a theoretical basis for scientifically evaluating the pollutant carrying capacity of Guanghai Bay. This comprehensive understanding of pollutant transport and fate will support more informed decision-making in environmental management and pollution control efforts, ensuring the protection and sustainability of the marine environment in Guanghai Bay.

The offshore nuclear power plant site adopts a secondary circulation cooling water supply method, equipped with large countercurrent seawater natural ventilation cooling towers. Due to the effect of cooling towers as tall buildings and the mist plume emitted from cooling towers on the diffusion of airborne radioactive substances, it is necessary to conduct research on the effect of large natural ventilation cooling towers on the diffusion of nuclear power plant sites based on the engineering layout. The effects of cooling towers and their plumes on the diffusion of airborne radioactive substances through wind tunnel experiments and numerical simulations were summarized in this study. According to the characteristics of wind distribution on representative nuclear power plant sites, in combination with the distribution of residential areas, the dominant wind direction E, the secondary dominant wind direction N, and the wind direction W from the cooling tower to the nuclear island were used as the experimental wind directions. The influence of cooling tower air inlet during the experiment was considered. The site is comprehensively affected by buildings and mountains, and the effect of buildings such as cooling towers on the site is dominant. No matter whether the thermal plume is discharged or not, the large natural ventilation cooling tower has entrainment effect, and the airborne pollutants discharged from the stack will enter the cooling tower through the air inlet at the bottom of the cooling tower and act as an elevated source for secondary discharge. The entrainment effect with hot plume discharge is more significant than that without hot plume discharge, and the concentration of pollutants discharged through the cooling tower significantly increases. The wind tunnel experiment results show that when the chimney is discharged at a height of 75 meters, the entrainment fraction of the large natural ventilation cooling tower is 29.8% when it is working, and 1.2% when it is not working. The numerical simulation results show that when the chimney is discharged at a height of 75 meters, the entrainment fraction of the large natural ventilation cooling tower is 25.28% when it is working, and 0.75% when it is not working.

In order to scientifically assess the radiation risk in the offshore environment, it is necessary to deepen the research and understanding of the adsorption and transport behaviors of radionuclides in the offshore environment. In this paper the content and distribution characteristics of the major natural and artificial radionuclides 238U, 226Ra, 210Pb, 40K and 137Cs in the sediments of the Northern Bohai Sea were researched, and the key factors affecting them were analyzed. After systematically organizing the data on radionuclide contents of sediments and seawater in the Chinese offshore, the migration and diffusion capacities of 238U, 232Th, 226Ra, 40K and 137Cs in different areas of the Chinese offshore were quantified by analyzing sediment distribution coefficients (Kd, the lower of Kd, the greater the capacity for migration and diffusion), and then the radiation risks of 238U, 232Th, 226Ra, 210Pb and 137Cs to pelagic and benthic fish in the Chinese coastal waters were preliminarily evaluated by using ERICA software based on the statistical method of measured data and built-in parameter simulation method. The results show that sediment grain size and organic carbon content are the key factors affecting the adsorption/desorption process of radionuclides in sediments. Under the influence of physical/chemical adsorption, both sediment particle size and organic carbon content are well correlated with radionuclide content, and the correlation is particularly significant for 210Pb, which has a high particle activity. Different radionuclides have different migration and diffusion capabilities of Chinese offshore (40K (Kd=7.8×102 L/kg)>137Cs (Kd=9.8×102 L/kg)>226Ra (Kd=2.3×103 L/kg)>232Th (Kd=4.3×103 L/kg)), there are also differences in the migration and diffusion behavior of the same nuclide in different sea areas (for 226Ra, Bohai Sea (Kd=1.1×103 L/kg)>Yellow Sea (Kd=2.9×103 L/kg)>South China Sea (Kd=3.1×103 L/kg)>East China Sea (Kd=4.4×103 L/kg), for 137Cs, Bohai Sea (Kd=4.7×102 L/kg)>South China Sea (Kd=7.9×102 L/kg)>Yellow Sea (Kd=1.5×103 L/kg)), although the overall radiation risk to Chinese offshore fish is low (total risk quotient

This paper presents a novel artificial intelligence (AI) based approach to predict crucial meteorological parameters such as temperature, pressure, and wind speed, typically calculated from computationally intensive weather research and forecasting (WRF) model. Accurate meteorological data is indispensable for simulating the release of radioactive effluents, especially in dispersion modeling for nuclear emergency decision support systems. Simulation of meteorological conditions during nuclear emergencies using the conventional WRF model is very complex and time-consuming. Therefore, a new artificial neural network (ANN) based technique was proposed as a viable alternative for meteorological prediction. A multi-input multi-output neural network was trained using historical site-specific meteorological data to forecast the meteorological parameters. Comprehensive evaluation of this technique was conducted to test its performance in forecasting various parameters including atmospheric pressure, temperature, and wind speed components in both East-West and North-South directions. The performance of developed network was evaluated on an unknown dataset, and acquired results are within the acceptable range for all meteorological parameters. Results show that ANNs possess the capability to forecast meteorological parameters, such as temperature and pressure, at multiple spatial locations within a grid with high accuracy, utilizing input data from a single station. However, accuracy is slightly compromised when predicting wind speed components. Root mean square error (RMSE) was utilized to report the accuracy of predicted results, with values of 1.453 ℃ for temperature, 77 Pa for predicted pressure, 1.058 m/s for the wind speed of U-component and 0.959 m/s for the wind speed of V-component. In conclusion, this approach offers a precise, efficient, and well-informed method for administrative decision-making during nuclear emergencies.

Phytoremediation is an environmentally friendly and highly promising method for soil remediation, with the core issue being the selection of suitable plants and applicable conditions. During the remediation of uranium-contaminated environments, the enrichment of uranium adversely affects the normal growth of plants, causing stress effects. This study aimed to explore the stress response mechanism of Bidens pilosa L. to uranium in the soil and the influence of chelating agents on its uranium enrichment capacity through experiments on the remediation of uranium-contaminated soil. The phenotypic, physiological, and chlorophyll fluorescence parameters of Bidens pilosa L., as well as its uranium enrichment capacity, were studied under the influence of chelating agents and uranium. The results show that with the increase of uranium concentration, the levels of soluble protein, malondialdehyde (MDA), peroxidase (POD), and catalase (CAT) in Bidens pilosa L. first increase and then decrease. The maximum quantum yield (Fv/Fm) and non-photochemical quenching coefficient (NPQ) significantly decrease with the increase of uranium concentration, while the actual quantum yield (ΦPSⅡ) gradually increases. The application of chelating agents significantly enhances the uranium enrichment capacity of Bidens pilosa L. Low concentrations of chelating agents promote the synthesis of soluble protein, POD, and CAT, and reduce MDA content. Additionally, low concentrations of chelating agents increase the Fv/Fm and ΦPSⅡ values of Bidens pilosa L. However, they could also affect the open center structure of PSⅡ in plants, leading to the damage of the photoprotection system. High concentrations of chelating agents could directly cause plant death. In terms of enrichment capacity, when the soil uranium concentration is 50 mg/kg, under the action of 20 mmol/kg oxalic acid, Bidens pilosa L. achieves a maximum total enrichment coefficient of 6.29, with an above-ground part enrichment coefficient of 4.26 and a transfer coefficient of 2.09, exceeding the lower limits defined for hyperaccumulator plants. Thus, it can be defined as a hyperaccumulator plant, but under these conditions, Bidens pilosa L. does not survive for more than 24 h. Therefore, if Bidens pilosa L. is used as a phytoremediation material for uranium-contaminated soil, the application of 10 mmol/kg citric acid can enhance its uranium enrichment capacity while ensuring better survival rates. The above results indicate that by adjusting the concentration of chelating agents, it is possible to effectively increase the uranium enrichment efficiency of Bidens pilosa L. while minimizing damage to the plant itself, thereby playing a more significant role in the remediation of uranium-contaminated soils. Furthermore, understanding the precise biochemical and physiological responses of Bidens pilosa L. to uranium stress and chelating agents can help in developing more refined strategies for using this plant in various contaminated environments. This deeper insight into the plant’s response mechanisms also opens up avenues for interventions to enhance its resilience and accumulation capacity, making phytoremediation a more viable and efficient option for large-scale environmental cleanup efforts.

Radioactive nuclides such as 129I are produced during the operation of nuclear facilities. 129I is volatile and can induce radiation sickness. Capture of radioactive iodine can protect public health and facilitate the sustainable development of the nuclear industry. Preliminary studies on the synthesis and iodine adsorption properties of thorium based MOFs have shown that thorium-based MOFs usually have different structures and iodine adsorption performance when they contain different ligands. Quinoline derivatives containing both hydroxyl and carboxyl groups are a class of rigid ligands with strong coordination ability. Although these ligands have been used to synthesize transition metal complexes, their application in the preparation of thorium MOFs remains unexplored. To further investigate the effect of ligands on the structure and iodine adsorption performance of thorium-based MOFs, in this paper thorium-quinoline MOF (Th-HQCA) was synthesized using 2-hydroxyquinoline-4-carboxylic acid (HQCA) as the ligand via solvothermal method, and its structure, composition, thermal stability, spectral properties, and iodine adsorption performance were investigated through various characterization techniques, including single-crystal X-ray diffraction, powder X-ray diffraction (PXRD), scanning electron microscope-energy dispersive X-ray spectroscopy (SEM-EDS), Fourier transform infrared spectroscopy (FT-IR), ultraviolet-visible (UV-Vis) spectroscopy, thermogravimetric analysis (TGA), X-ray photoelectron spectroscopy (XPS), and fluorescence spectroscopy. Crystallographic analysis reveals that the coordination modes of thorium ions in the structure of Th-HQCA are the same, and each ion is coordinated with four carboxylate oxygen atoms from four HQCA and four hydroxyl oxygen atoms from another four HQCA. It is worth mentioning that thorium ions usually coordinate with carboxylate oxygen atoms in ligands, and rarely coordinate with hydroxyl group. The connections between thorium ions and HQCA ligands result in the formation of a novel 2,8-connected framework structure of Th-HQCA. The channels of the framework are filled with water molecules. Based on the crystallographic analysis, the ratio of thorium ions, HQCA, and water molecules is determined to be 1∶4∶1. However, further thermogravimetric analyses indicate that there may be twelve cocrystallized water molecules in the compound. Consequently, the chemical formula for Th-HQCA can be expressed as [Th(C9H5NOCOO)4]·12H2O. The solid-state fluorescence spectrum shows that Th-HQCA crystals emit blue light when they are excited by ultraviolet light. The adsorption experiments of Th-HQCA on gaseous iodine and iodine in cyclohexane solution show that their adsorption kinetic curves fit the pseudo-second-order kinetic equation, and the adsorption isotherm of the latter fits the Langmuir model. Through further calculations, the maximum adsorption capacity of Th-HQCA for iodine vapor is 535 mg/g, and the maximum adsorption capacity for iodine in cyclohexane solution is 295.63 mg/g. In summary, Th-HQCA is a thorium-quinoline MOF with a novel 2,8-connected topological structure, and it shows remarkable fluorescence properties and can effectively capture iodine in the gas phase and solution.

Comisión Nacional de Energía Atómica (CNEA) has the responsibility for restoring uranium mining facilities once the operations have finished. CNEA, within its Environmental Program and in compliance with its legal responsibilities, decides to implement a restoration project for all sites related to the mining and processing of uranium ores. The Malargüe Site is located within the Province of Mendoza in the city of Malargüe. It is the first site to successfully complete its remediation. The activities consist of relocation of tailings to an engineering repository. The tailings management (encapsulation) and rehabilitation of the area was finished in June 2017. The remediation alternative for the ore tailings was selected after conducting comparative studies and submitted the project to the society for consideration. The objective of the encapsulation of the mineral tails is to isolate them from the environment, also proceeding with the decontamination and rehabilitation of the area (landscaping, post-closure monitoring and 20 years monitoring period). Encapsulation consisted of the construction of a containment cell for the mine tailings, to isolate them and prevent pollutants from entering the environment through the transfer routes. To clean the impacted areas, the soil was removed, it was incorporated into the encapsulation, and the filling was carried out with natural soils from the area. Remediation prevents radon transfer to the environment, as 222Ra is an alpha emitter with a half-life of four days, which produces its own radioactive progeny. Radon progeny are solids, and when a 222Ra nucleus emits an alpha particle into the air, the resulting 218Po nucleus, momentarily electrically charged, adheres to any dust particle. Remediation prevents the discharge into the air containing radon and also containing dust particles charged with intensely radioactive radon progeny. The tasks mentioned make it possible to decrease radon emanation, reduce radiological risks to the public and prevent the entry of rainwater into the system. In addition, the containment system prevents the discharge of contaminated liquids into the environment, avoiding contamination of the groundwater. All these activities are according to the concepts of sustainability.

Due to its advantages of higher heat flux and wider operating temperature range, sodium heat pipes have shown promising application prospects in related miniaturized and unmanned nuclear reactor platforms. The accuracy of the working fluid’s evaporation accommodation coefficient (acc) in sodium heat pipes holds significant importance for their thermal transport analysis. According to existing research, it has been found that the evaporation characteristics of the working fluid are related to the contact surface condition, and the value of acc is significantly influenced by temperature. Therefore, the impact of different internal parameters and operating temperatures on acc of sodium within the sodium heat pipe was studied by using experimental methods. By varying parameters such as heating power, inclination angle, filling ratio, and capillary wick, the axial outer wall temperature and internal vapor-phase temperature under steady-state conditions were measured using the thermocouple placed on the outer wall and the radially movable thermocouple inside the heat pipe. The experimental results reveal that the vapor-phase temperature distribution is non-uniform in the radial direction on steady-state condition, exhibiting a certain thermal gradient. Different methods for selecting vapor temperatures, in combination with measured outer wall temperature, were employed to derive the experimental values of acc under different work conditions and operating temperatures based on the Hertz-Knudsen-Schrage equation. Analysis demonstrates a good linear relationship between the logarithmic value of acc and the logarithmic saturation pressure within the sodium heat pipe, which remains consistent for different sodium contents and inclination angles. However, when using the wall temperature of the adiabatic section or the vapor-phase temperature of the adiabatic section as the vapor temperature, acc tends to be relatively small due to the influence of radial thermal resistance and the wetting properties of liquid sodium, particularly below the wetting transition temperature (550 ℃), but it still satisfied a logarithmic linear relationship. Nevertheless, as the contact angle between the liquid sodium and the capillary wick approached 0°, the effect of the capillary wick became more pronounced, leading to a significant increase in acc due to the axial spreading facilitated by the excellent wetting properties of liquid sodium, therefore, the value of acc after the wetting contact angle transition temperature is also relatively larger. Overall, there is still a good linear relationship between the logarithmic value of acc and the logarithmic saturated vapor pressure. Finally, through fitting analysis, an empirical relationship for acc of sodium within the heat pipe at different temperatures was established, which can be used to predict acc under various operating conditions.

The compact design of the marine nuclear power reactor results in a significant reduction in the size of the main equipment, particularly the steam generator. The primary steam-water separator is the core component in this regard. The swirl vane separator is the primary separator in the steam-water separator, responsible for more than 80% of the steam-water separation task. The performance of the marine nuclear power system directly affects its safety and economy. The primary focus of this study centered on the swirl vane separator, employing the Euler-Euler two-phase flow model in conjunction with the RNG k-ε turbulent flow model. Furthermore, an additional inertia force resulting from the rolling motion was integrated into the momentum equation as a source term using user-defined functions (UDF). The impact of varying gravity components was considered, leading to the development of a three-dimensional numerical computational model to analyze the flow and separation of steam and water within the separator. The impact of varying rolling angle and rolling period on the flow properties and operational efficiency of the separator was methodically investigated. The results indicate that the turbulent mixing of the steam-water two-phase fluid within the separator is enhanced by the rolling motion, leading to a chaotic pressure field, velocity field, and distribution of liquid volume fraction. The critical state is observed when the rolling angle reaches 40° and the rolling period is 2 s. This critical state causes fluid backflow in different parts of the separator, greatly affecting the effectiveness of separating the two-phase flow. The additional inertia force from the rolling motion results in periodic pressure and velocity fluctuations within the separator, with a phase delay of one-quarter period compared to the separator’s motion. In the near-wall region of the separation cylinder, the liquid volume fraction reaches its peak when the separator is at its maximum inclination angle. The periodic changes in separation efficiency follow a sinusoidal pattern. Before reaching the critical condition, the variation in separation efficiency corresponds with the rolling motion of the separator. The influence on the separation performance becomes more pronounced with higher rolling magnitude and shorter rolling period.

Oxygen is the most promising non-metallic inhibitor in lead-bismuth cooled fast reactors (LFRs). The addition of oxygen to the coolant can format a protective oxide layer on the surface of structural materials, which will effectively alleviate the corrosion of structural materials by the liquid lead-bismuth eutectic (LBE). LFR is a complex environment characterized by the interaction of multiple physical fields. For instance, the growth and removal behaviors of the oxide layer are influenced by various factors such as temperature, oxygen concentration, coolant velocity, and time. Moreover, the formation of the oxide layer changes the thermal-hydraulic characteristics and neutronics parameters of the reactor core. Therefore, studying the coupled effects of oxidation corrosion, thermal-hydraulics, and neutronics is of paramount importance for the development, design, and safety assessment of LFRs. A multi-physics framework that couples neutron physics, thermal-hydraulics, and material corrosion was proposed to investigate the multi-physics coupling characteristics of the fuel assembly in LFRs. Within the coupling framework, neutronics calculations were performed using the open-source neutron diffusion equation solver Moltres, thermal-hydraulic calculations were conducted using the Navier-Stokes module included in the multi-physics object-oriented simulation environment (MOOSE) platform, and corrosion calculations were carried out using the Seal module developed based on the MOOSE. The coupling framework involves two types of coupling parameter transfer relationships: 1） The oxidation corrosion field obtains coolant temperature and flow velocity from the thermal-hydraulic field to compute the oxide layer thickness and transfers the oxide layer thickness to the thermal-hydraulic field to calculate the convective heat transfer coefficient; 2） The neutron physics field receives temperature distribution from the thermal-hydraulic field to compute keff, neutron flux distribution and power distribution, and transfers the power distribution to the thermal-hydraulic field for thermal-hydraulic calculations. In terms of numerical system solving, the coupling framework employs the concept of directly coupled equations of the three physical fields and solves them using the Newton-Krylov iteration method. A neutronics-thermal-hydraulics-material coupling problem of a 19-rod bundle fuel assembly in an LFR was computed using the coupling framework, and the effects of oxygen concentration and coolant inlet temperature on the temporal variations of key coupling parameters and the distribution of oxide layers were investigated. The results indicate that under the benchmark condition, after 10 000 hours of oxidation corrosion, the average thickness of the oxide layer on the fuel assembly cladding surface is approximately 9.86 μm. The maximum temperature rises of the fuel and cladding are 13.36 K and 5.63 K, respectively, with a decrease in keff of 7 pcm. Increasing oxygen concentration is beneficial for inhibiting magnetite dissolution and enhancing the self-repair ability of the oxide layer, but the promotion effect of increasing oxygen concentration on the growth of Fe-Cr spinel is limited after reaching a certain concentration. Although raising the coolant inlet temperature leads to an increase removal rate of magnetite on the inner surface of the cladding at the center of the assembly, it significantly promotes the growth of Fe-Cr spinel.

Uncertainties on results of reactor physics calculations basically originate from uncertainties of solvers, modeling parameters and nuclear data. The uncertainty quantification (UQ) of import core parameters is critical for the safety and reliability of the innovative nuclear reactor designs. In this paper, the UQ of the keff due to the neutron cross section data and the manufacturing tolerance of modeling parameters for a small prismatic high temperature gas-cooled reactor (HTGR) was reported. The stochastic sampling (SS) method, the sensitivity analysis (SA) method and the polynomial chaos expansion (PCE) method have been adopted during the UQ process. Firstly, the sensitivity analysis of keff to nuclear data was performed and uncertainties were calculated with sensitivity coefficients and covariance matrix of cross sections. The numerical results show that the keff uncertainty due to the neutron cross section data is about 511 pcm, among which 311 pcm are introduced by the neutron cross sections of the graphite and 28Si, the considerable fraction of contribution compared to the conventional 235U and 238U. In addition, the SS method and PCE method were used to assess keff uncertainties owing to 25 modeling parameters, including important geometry dimensions and material compositions. A large number of input parameters were sampled and physical calculations were repeatedly conducted for each set of sampled parameters. In the PCE analysis, Legendre basis was used to construct polynomials on the assumption that all modeling parameters followed a uniform distribution and the linear regression method was adopted to solve for the coefficients. The analysis results indicate that the control of the uranium loading is the most critical manufacturing requirement. Although the uncertainties of the sizes of TRISO particles are large due to the manufacturing capacity, as long as the total loading of uranium is chosen as an individual control parameter and decoupled from other modeling parameters, keff uncertainty introduced by 25 modeling parameters can be reduced from about 1 950 pcm to about 420 pcm, among which the thickness of carbon layers of TRISO particles, the total loading and the enrichment of uranium have the most significant contributions. As far as the UQ analysis methods are concerned, the PCE and the SA methods show better efficiency than the SS method, and they can produce the total uncertainty and the individual uncertainty due to each parameter simultaneously. In addition, the PCE method gives a surrogate model at the same time to predict the keff for different modeling parameters without resolving to the time-consuming design calculations.

The exploration of flow-induced vibration in fuel rods is of utmost importance as it plays a pivotal role in comprehending and mitigating factors that contribute to fuel failure and reactor shutdown. This understanding is crucial for advancing the nuclear energy industry. To unravel the intricacies of this complex phenomenon, a sophisticated high-fidelity finite element model of fuel rods was meticulously constructed. This model serves as the cornerstone for a computational analysis of flow-induced vibration responses utilizing ANSYS Batch, firmly roots in the principles of random vibration theory. In the pursuit of a profound understanding, support stiffness values to simulate and scrutinize diverse scenarios that fuel rods might encounter were systematically varied. The outcomes of these simulations were meticulously compiled to establish a comprehensive database of flow-induced vibration responses. This database stands as a valuable resource for future research endeavors and engineering applications within the nuclear energy domain. Taking a step forward, an innovative approach was adopted to streamline the analysis process. Leveraging the snapshot matrix derived from the extensive database, a high-fidelity reduced-order model (ROM) was developed. Two data-driven methodologies, namely proper orthogonal decomposition (POD) and dynamic mode decomposition (DMD) methods, were employed to construct the ROM. This ROM enables the rapid reconstruction of flow-induced vibration responses, facilitating efficient analysis and decision-making in real-world applications. A critical facet of this paper involves a comparative analysis of the reconstruction effectiveness of fuel rod flow-induced vibration responses using both the POD and DMD methods. The results of this comparison reveal that, when considering the reconstruction of vibration responses with the same number of modes, the POD method outperforms the DMD method. This finding underscores the importance of selecting appropriate methodologies based on specific objectives and computational efficiency. Furthermore, the results indicate that the DMD method excels not only in efficiently reconstructing the vibration responses of fuel rods but also offers the unique capability to assess the stability of each DMD mode. This dual functionality enhances the overall diagnostic capabilities of the DMD method, providing valuable insights into the dynamic behavior and potential instabilities of the fuel rod system. In conclusion, the exhaustive investigation outlined in this paper not only significantly contributes to the understanding of flow-induced vibration characteristics in fuel rods but also provides a robust framework for developing advanced models and methodologies for future research and practical applications in the nuclear energy industry. The comprehensive nature of our approach ensures that our findings are not only insightful but also applicable in shaping the future of nuclear energy technology.

In a speech on 2 June 2023, Jake Sullivan, the US National Security Adviser, expressed willingness to engage in nuclear arms control discussions with Russia “without preconditions”. Although the offer was rejected by Moscow, the White House initiative provides with a “textbook case” for the theory of negotiation. Through the use of the Harvard Negotiation Method, this paper sets out the theoretical interests of Russia and the United States in engaging in arms control discussion, in particular the explicit intention of both sides to prevent nuclear escalation and avoid an arms race. It then explains why, in reality, these interests diverge over the broader security landscape, including the US support for Ukraine. This contribution outlines US and Russia’s alternatives to negotiation, possible common ground in the area of “damage-limitation” efforts, as well as creative options for mutual gain. The authors ask why the US attempt to “compartmentalize” arms control, grounded on the Cold War tradition, has failed to convince Russia. They offer practical recommendations to facilitate constructive dialogue and progress in arms control discussions, building on lessons learned from the current “dialogue of the deaf”.

There is a substantial amount of experimental data that confirm the peculiarity of the oxide fuel behavior during the first hours after a reactor reaches its power. With an increase in temperature during the said period, oxide fuel pellets crack because of a significant temperature gradient. Further developments occur due to the accumulation and redistribution of fission products, and manifest themselves as changes in the fuel matrix porosity and the formation (or diameter increase) of a central hole. The zones that differ in their microstructure, density and thermal conductivity are formed along the fuel pellet radius. As the oxide fuel composition and restructuring result in its thermal conductivity change, it is important to pay attention to the formulas used in the stress-strain state calculations of a fuel element. The methodology for calculating changes in the oxide fuel porosity and the pellet’s internal diameter is proposed and based on the published research papers dedicated to the study of oxide fuel properties and behavior during the first hours after a reactor reaches its power. Тhe methodology was tested using real experimental data on the porosity redistribution along the fuel pellet radius. The presence and the size of the pellet’s inner hole, as well as changes in the fuel matrix porosity, have a noticeable effect on the maximum temperature value. Taking into account the pellet structure evolution while performing computational simulation of the fuel rod operation under irradiation allows assessing the fuel element’s operability more accurately. The proposed methodology can be used in computer codes designed to calculate the stress-strain state of cylindrical fuel elements of fast reactors.

The paper presents an experimental evaluation of drying the damaged VVER SNF after the wet storage. The aim of the experiments was to justify the fire safety of air-tight canisters with the dried SNF during transportation, storage, and reprocessing. Keeping the SNF in boric acid during the experiments determined the concentration of uranium, plutonium, and specific activity of fission products in the solution inside the model canisters with the SNF. The canisters with the SNF underwent the thermal vacuum drying to determine the process parameters (temperature, residual pressure, drying rate), a hydrogen release, 85Kr activity during the drying, and an airborne radionuclide release onto the filters of the SNF thermal vacuum drying system. The results of weighing the baskets of the canisters before soaking in the boric acid and after drying at 160 °С to the residual pressure below 6 mbar showed that the difference in weight is almost the same as the calculated difference in weight of H3BO3 loaded into the canister and the weight after the complete decomposition into B2O3. Thus, we have discovered that the boric acid undergoes complete decomposition into boric anhydride, and the SNF corrosion products generate hydrogen that accumulates in the air-tight canister. We also considered the hydrogen that accumulated in the air-tight model canister with the dried SNF as a result of the radiolysis of hydrated corrosion products of the spent oxide nuclear fuel, as well as a release of gaseous fission products (GFP) inside the canister. We calculated the quantity of the hydrated corrosion products of uranium oxide after the wet storage and determined their chemical composition after the thermal-vacuum drying. The equation with an exponential decay function describes the hydrogen accumulation. The calculated maximum volume of hydrogen and uranium in the SNF corrosion products suggests that the SNF corrosion product is schoepite with the number of gyrated water molecules corresponding to the formulas from UO3·0.95H2O to UO3·1.5H2O. The received data allows for the fire safety justification of drying the wet damaged SNF and handling the dried SNF during transportation and storage.

The subject of constant support development is currently becoming increasingly important in connection with the national strategy of transition to a closed nuclear fuel cycle. At the same time, the importance of the tasks of refining calculation models and minimizing methodological, statistical, and constant errors is increasing. In this connection, the idea of creating a universal system of constants describing with equal accuracy both uranium loading and loading with mixed oxide uranium-plutonium fuel and the possibility to perform on its basis both precision Monte Carlo calculations (data format – ACE) and multigroup calculations (formats – ABBN and ANISN) was laid in the basis of this work. In this paper we analyze the freely available libraries of evaluated nuclear data in order to justify the choice of source files for the formation of a new system of group constants for neutron-physical calculations of fast reactor cores, describe the process of formation of a new library of reactor constants, and verify the obtained system on computational test models of BN-800 reactor and critical systems. The process of formation of the new system of group constants included selection of initial neutron data files, updating of data tables of basic neutron cross-sections, self-shielding factors and Doppler coefficients, as well as data on fission spectra for the main fuel nuclides. The methodological component of the error for test models of the BN-800 core was evaluated. Using the ABBN-RF22 system of group constants it was possible to estimate the correction in the 299 groups calculation, which amounted to 0.3%. Previously, when using the ABBN-93 library, there was no such possibility due to the lack of continuity between the files of evaluated neutron data and the group constants used.

Using strong electromagnetic fields generated by lasers to interact with electrons for precise diagnosis and manipulation of electron beams represents a recent focal point in accelerator technology. This approach surpasses the limitations of conventional RF technology, such as low electric field gradients and timing jitters, effectively enhancing the accuracy of ultrafast electron beam diagnostics and manipulations. As demands for precision continue to rise, the precise diagnosis of crucial parameters of ultrafast electron beams remains challenging. This study delves into the electromagnetic behavior of THz-driven devices and proposes an all-optical method utilizing single-cycle THz radiation to compress and characterize a 3 MeV electron beam. Particle tracking simulations demonstrate an astonishing compression effect, reducing the bunch length from 54.0 fs to 4.3 fs, and achieving sub-femtosecond bunch length measurement resolution. Moreover, when combined with an orthogonal THz streak camera, this method shows even greater potential in multi-bunch scenarios.

Nuclear energy is used more and more widely as a clean energy source, but nuclear energy facilities are risky, and when a nuclear leak occurs, it is necessary to detect equipment in a nuclear radiation environment. In the nuclear radiation environment, due to the impact of high-energy particles on the camera sensor, the collected image contains a lot of radiation noise, which greatly reduces the visual perception of the image. Aiming at the problem that radiation noise reduces image quality, a radiation image compound filtering algorithm combining median filtering and multi-frame average filtering is proposed based on radiation noise characteristics. Compared with several common filtering algorithms, the radiation noise image is filtered under the same radiation dose and achieve the highest Peak Signal Noise Rate (PSNR) and Structural Similarity (SSIM), and compared with the multi-frame average filtering method, the number of image frames required by the algorithm in this paper is greatly reduced. Experimental results show that the algorithm can effectively eliminate radiation noise and is more suitable for image filtering in radiation environment.

The binary decision diagram (BDD) method provides a means to calculate the exact probability of a fault tree, but cannot solve large fault trees for nuclear power plant PSAs. However, the BDD method can be used as a supplementary means to increase the accuracy of PSA quantification, especially for a seismic PSA which includes many non-rare events.Since it is difficult to solve a large PSA model using BDD method, several approaches have been developed that partially apply the BDD technique to small-sized branches of a large PSA model. This paper proposes an approach of partially applying BDD technique to large-sized branches. The results from the pilot application to a seismic PSA model shows that the partial BDD approach of this paper provides a relatively accurate way to quantify each sequence as well as the overall core damage in a seismic PSA model.

Preventive Maintenance(PM) for safety component during power operation at nuclear power plants, On-Line Maintenance(OLM) refers to intentionally entering the Limited Condition of Operation(LCO) specified in the Technical Specification(TS) for safety-related systems and components in order to perform preventive maintenance within the Allowed Outage Time (AOT). This study assessed the feasibility of conducting OLM at the domestic APR1400 nuclear power plant. It focused on preventive maintenance duration and risk perspectives. A total of 78 FEGs were developed for 4450 facilities, considering system functions and preventive maintenance scope during output operation for eight safety-related systems. Additionally, maintenance items included in FEGs were selected, designated as targets for OLM, and their maintenance durations were evaluated and compared with AOT for each maintenance item. As a result, the Auxiliary Feedwater and Essential Chilled Water systems were identified as systems allowing OLM. Furthermore, utilizing the Risk Monitoring System (RIMS), the increased risk value due to the unavailability of target equipment during preventive maintenance was analyzed to determine whether it falls within the acceptable range.Regarding the temporary risk increase caused by OLM, it was observed that in all systems, it falls within Zone III according to NUMARC93-01 standards, allowing for normal equipment arrangement for OLM. However, according to the risk increase standards rate in domestic nuclear power plants, when maintaining the A-train in four systems including Component Cooling Water, they are all evaluated as 'Orange,' indicating that measures for risk mitigation are necessary for OLM to be feasible. When considering extending AOT up to 1.6 times the maintenance time, the risk increase falls within Zone III according to permissible change in risk standards, indicating that AOT extension might be feasible based solely on risk changes. To apply OLM within the permissible risk management scope in domestic nuclear power plants, regulatory policies need to allow voluntary LCO entry for preventive maintenance, necessitating clear determination by regulatory agencies using risk-informed policies. While OLM seems viable concerning maintenance duration and quantitative risk aspects, for inducing regulatory policy changes, comprehensive OLM guidelines are necessary, including risk management strategies.

The small modular PWR (SMPWR) usually adopts integral design. Under severe accident, the system responses are different from those large PWRs. It is necessary to study the severe accident behavior of the SMPWR. A MELCOR model is developed for SMPWR and its steady-state results are in good agreement with the design values. Severe accidents induced by secondary pipeline break accidents are simulated, and no pressure relief measures are taken to keep the primary loop under high pressure. The mitigation effects of passive containment air cooling system (PAS) and passive cavity injection system (PCIS) are evaluated under different cases. The results show that under high pressure conditions, PCIS can effectively cool the lower head. The earlier the PCIS operates, the more significant the mitigation effect can be. In addition, PAS can effectively reduce the peak pressure and temperature in the containment. This study can provide a reference for the formulation of severe accident management guidelines on SMPWRs.

The Fukushima accident showed that the safety of multiple nuclear power plants (NPPs) at the same site could be jeopardized simultaneously. Since then, many studies have focused on developing strategies to prevent the spread of multi-unit accidents, with numerous countries establishing strategies to use mobile equipment. However, mobile equipment strategies are inherently accompanied by a high degree of uncertainty regarding operation success and duration because multiple organizations and personnel interact in various ways during multi-unit accident situations. Furthermore, supplementing current fixed equipment with additional mobile equipment requires extra resources. Therefore, cross-tie strategies that use currently installed fixed equipment can provide additional means to manage site risk with relatively few additional costs. This study proposes a multi-unit probabilistic safety assessment–based risk evaluation framework for utilizing cross-tie systems in NPPs and a modeling methodology to quantify the effectiveness of the cross-tie strategies. A case study was conducted to evaluate the risk reduction from using cross-tie strategies for emergency diesel generators and alternate AC diesel generators, which are power systems utilized in multi-unit loss of offsite power initiating events. It is expected that the developed framework and methodology can be utilized for other types of cross-tie strategies as well.