BackgroundThe 1.2 MeV/10 mA electron accelerator, as one of electron irradiation sources for comprehensive irradiation test chamber in Space Environment Simulation and Research Infrastructure (SESRI), can provide electron beams on the millimeter scale. However, the electron beam, as the electron irradiation source in space environment ground simulation experiment for aerospace, must be uniformly irradiated on large objects. Therefore, a well-suited irradiated technique is significant.PurposeThis study aims to obtain beam scanning with less than 10% of inhomogeneity for 0.6 MeV, 1.0 MeV, 1.2 MeV electron beam for SESRI.MethodsBased on the overall irradiation requirements, a specific beam-scanning system, including the scanning magnet with customized design, digital power supply and a dedicated apparatus for beam uniformity measurement, was developed. Particularly, in order to eliminate the influence of 45° incidence of electron beam upon scanning uniformity, an asymmetric and non-standard triangular waveform for the magnetic field excitation current was employed and implemented. The measurements for the beam non-uniformity were carried out on field experiments.ResultsExperimental results show that the scanning area of this electron accelerator reaches 1 000 mm×1 000 mm, and the scanning nonuniformity is less than 10% for variable beam energy from 0.6 MeV to 1.2 MeV, achieving the design goal and satisfying the irradiation requirements of SESRI.ConclusionsA specific beam-scanning system developed for SESRI is verified in this study, offering a good reference for any similar beam-scanning scenarios.

BackgroundThe use of controlled X-ray sources instead of 137Cs radioactive sources in density logging has become a new trend. The intensity of the X-ray source is substantially influenced by the high voltage on the target substrate, and the density measurement uncertainty can be maintained at 0.01 g·cm-3 when the high voltage is 350 kV.PurposeThis study aims to analyze the parameters of the shielding material and thickness suitable for the 350 kV high-voltage X-ray density logging instrument.MethodsThe Monte Carlo method was used to analyze the energy spectrum and counting rate of X-rays passing through different materials and thicknesses. By comparing the correlation between the 0~0.15 MeV and 0.15~0.35 MeV energy windows, the reasons for the difference between the X-ray attenuation and detector count rate in different energy windows were determined. In addition, combined with the actual instrument model construction of the four-detector X-ray density logging instrument, the influence of the three parts of particles on the detector was primarily considered. The placement mode and optimal thickness of each part of the shield for detectors were analyzed and designed using Monte Carlo N-particle (MCNP) simulation.ResultsThe simulation results show that the attenuation of X-rays in high- and low-energy windows increases with increase of atomic number and thickness of shielding materials. When tungsten nickel iron alloy is selected as the shielding material for the four-detector X-ray density logging instrument model, the suitable thickness of the shield between the base and the near-source detector is 1.75 cm. Meanwhile, to maintain the high voltage of X-ray generator at 350 kV, a shield layer with a thickness of 0.2 cm is placed between each detector, and a shield layer with a thickness of 0.35 cm is added to the back of the detector.ConclusionsThis study provides the design theory and key parameters for shielding materials and structures in the development of X-ray density logging tool.

BackgroundThe defects generated during the working process of metal materials have a significant impact on their performance. For example, the radiation-induced embrittlement and hardening of reactor pressure vessel (RPV) steels are a factor of concern, which hinders the life extension of the RPV. Annealing treatment is applied to alleviating irradiation-induced precipitates and defects and recover RPV's mechanical properties in the past few decades to extend the in-service lifetime of the RPV. Unfortunately, this conventional method generally requires a high treatment temperature and long operation time, inevitably wasting considerable energy due to the huge size of the RPV. Recently, as a more convenient and energy-saving method, the repair of metal defects by electropulsing treatment (EPT) has been developed.PurposeThis study aims to design and construct a device for EPT processing of samples, and investigate the repairs of defects in electron irradiated and deformed iron and RPV steel after EPT by using positron lifetime spectroscopy.MethodsElectron irradiated pure iron and RPV steel samples were prepared and subjected to multi parameter EPT device developed in laboratory, and the changes in defects of the samples with EPT were characterized by positron lifetime spectroscopy. In addition, the mechanical properties of pure iron tensile samples were characterized by micro Vickers hardness, and the defect information was characterized by positron lifetime spectroscopy to explore the relationship between macroscopic properties and microstructure.ResultsThe defects introduced by electron irradiation in pure iron and RPV steel samples gradually recover after EPT and exhibit similar patterns to annealing treatment. After stretching, the number of defects in pure iron samples increases, leading to an increase in Vickers hardness. EPT can restore defect and reduce Vickers hardness.ConclusionsThe defects generated by irradiation or deformation in pure iron and RPV steel can be partially repaired through EPT. The effect of defect repair is not only related to the initial state of the sample, but also to EPT's parameters. As a new non-destructive testing method, positron annihilation is expected to provide a criterion for material damage or defect repair under the action of pulse current, which can conveniently, quickly, and sensitively detect the defect state of actual working components.

BackgroundIn uranium mining and smelting, a significant amount of uranium-containing wastewater is generated, which can easily cause groundwater pollution. In polluted groundwater, uranium often combines with carbonates to form more diffusive and migratory uranyl carbonate complexes, making uranium removal more difficult.PurposeThis study aims to consider the performance of a three-dimensional electrochemical system of rGO/BB hydrogel particle electrodes on uranyl carbonate, and implement an efficient method for removing uranium-containing solutions.MethodsFirstly, graphene oxide (go) samples were prepared by modified Hummers method, then, the effects of electrolyte (sodium nitrate) concentration, applied voltage, electrode spacing, pH value, particle electrode dosage, and initial uranium concentration on the adsorption performance of uranyl carbonate ions by a three-dimensional electrochemical system were investigated, and the recycling performance of the system was studied. Finally, the adsorption mechanism of uranium was analyzed by scanning electron microscopy (SEM), Fourier transform infrared spectroscopy (FT-IR), Raman spectroscopy, and X-ray photoelectron spectroscopy (XPS).ResultsExperimental results show that under the conditions of pH value of 4~8, electrode spacing of 4 cm, voltage of 5 V, and rGO/BB hydrogel dosage of 90 mg, 200 mL of uranyl carbonate with a concentration of 1~1 000 mg?L-1 exhibits good adsorption efficiency. Even when the uranyl carbonate concentration is 1 000 mg?L-1, the adsorption rate can reach 87.56% in 14 h. After five cycles of adsorption and desorption, the adsorption rate remains above 87%, thus showing good recycling performance. The addition and use of rGO/BB hydrogel particle electrodes significantly improves the adsorption capacity of the electrochemical system for uranium, in which carboxyl and hydroxyl groups play a major role in uranium adsorption.ConclusionsThe results of this study show that the three-dimensional electrochemical system based on rGO/BB hydrogel particle electrodes has great application potential for removing uranyl carbonate ions.

BackgroundAviation γ spectrometer, measuring at high-altitude with low background counting rate, is prone to spectral line drift caused by factors such as long stable spectral periods and significant temperature effects due to its low background counting rate, hence seriously affects the accuracy of measurement results.PurposeThis study aims to propose an adaptive derivative-Gaussian joint peak search algorithm so as to quickly find and stabilize the peak value of airborne γ-ray spectrometer in low background environment.MethodsFirstly, the process of derivative peak finding algorithm was optimized so that it automatically determined the peak position, channel address, and left and right boundaries based on limited conditions, achieving adaptive transformation background window width deduction of Gaussian peak finding algorithm. Then, a self-developed miniaturized aviation γ spectrometer was used for flight measurement experiments by mounting it on unmanned aerial vehicle, and temperature changing measurement experiments by placing it in a variable temperature oven. Finally, the adaptive derivative gaussian peak finding results were compared and verified with that of traditional derivative peak-seeking algorithm and traditional gaussian peak-seeking algorithm, to ensure the accuracy of the proposed peak finding results and the efficiency of γ spectral stabilization.ResultsThe actual measurement results show that the adaptive derivative Gaussian joint peak finding algorithm has fast calculation speed and high accuracy. With 1 024 aviation tracks γ, the maximum spectral drift of the 40K peak measured by the energy spectrometer within the range of -20 ℃ to 50 ℃, does not exceed ±3 channels.ConclusionsThis study provides a new spectral stabilization method for the accurate measurement of airborne γ-ray spectrum in low background environment.

BackgroundIn recent years, the rapid advancement of nuclear energy and technology has led to the expanded utilization of special nuclear materials in more applications. In the pursuit of enhancing nuclear materials monitoring capability, extensive research has been dedicated to fast neutron multiplicity measurement systems based on liquid scintillation detector. However, these detectors are predominantly manufactured by foreign companies, such as Eljen and Saint-Gobain.PurposeThis study aims to evaluate the performance of a self-developed neutron detector based on EJ301 liquid scintillator.MethodsFirst of all, a 7.62 cm-diameter, 5.08 cm-thick liquid scintillator was applied to the development of neutron detector named EJ301-Z liquid scintillator detector, and 22Na and 60Co gamma sources were employed to calibrate this detector for establishing the relationship between incident particle energy and deposited energy. Subsequent tests on the neutron-gamma discrimination performance of the detector were conducted with a 252Cf neutron source, and figure-of-merit (FOM) values at different energy thresholds were calculated based on the charge comparison method. Then, comparison of the neutron-gamma discrimination performance of the detector with Eljen's EJ301 and Saint-Gobain's BC501A liquid scintillation detectors were conducted to evaluate the detector's performance relative to these commercially available models. Finally, the results obtained using the time-of-flight method were used to validate the results of the charge comparison method, providing an assessment of the absolute neutron-gamma discrimination performance of the liquid scintillator detector.ResultsThe experimental results show that, the EJ301-Z detect outperforms both Eljen EJ301 and Saint-Gobain BC501A detectors in terms of neutron-gamma discrimination, with an energy threshold of 150 keV, the rate of gamma mis-discrimination (RGMD) of EJ301-Z detector with the charge integration method is as low as 0.1‰.ConclusionsThe developed detector of this study demonstrates superior neutron-gamma discrimination performance compared to commercially available models from Eljen and Saint-Gobain.

BackgroundThe output signal quality of optical encoder in nuclear radiation environment is degraded due to the total ionizing dose (TID) effect of γ radiation.PurposeThis study aims to propose an improved adaptive line enhancer method (IALEM) that considers accuracy and efficiency to minimize the degradation of the output signal quality of photoelectric encoders in nuclear radiation environments caused by TID effect of γ radiation.MethodsFirstly, the Softsign function was introduced into the least mean square (LMS) algorithm to establish the nonlinear relationship between the error and the step size. Then, rapid convergence and small steady-state errors were achieved by introducing the previous step size value in the step-size-updating formula. Furthermore, this improved adaptive line enhancer method (IALEM) was compared with similar algorithms in terms of four aspects: convergence speed, steady-state error, low signal-to-noise input, and computational volume. Finally, the proposed algorithm was implemented in a field programmable gate array (FPGA) chip and verified on a photoelectric encoder platform, and its filtering effect was experimentally validated in a cobalt-60-source γ-radiation environment.ResultsThe results show that the proposed algorithm yields a higher convergence speed, lower steady-state error, and better filtering effect than that of other algorithms for low signal-to-noise ratio signals with less computational effort. Uniformity error and orthogonality error of the photoelectric encoder output signal after filtering by IALEM are reduced by 17.6% and 8.0%, respectively.ConclusionsThe experimental results show that the proposed algorithm can effectively filter out the noise generated through γ radiation and improve the output signal quality of the photoelectric encoder.

BackgroundTokamak plasma disruption generates a runaway current carrying enormous amounts of energy that, if not suppressed, can cause severe damage to equipment.PurposeThis study aims to investigate the effects of injecting a deuterium-argon/neon gas mixture on a runaway current during plasma disruption.MethodsBased on the high plasma current discharge conditions of the HL-2M tokamak device in China, numerical simulations were conducted using a fluid model in the DREAM code. Variations of plasma parameters, such as plasma current (Ip), ohmic current (Iohm), runaway current and the ohmic electric field, with the injected deuterium-argon content and ratio during the disruption process were consistently simulated.ResultsResults show that injecting a deuterium-argon/neon gas mixture suppresses the eventual formation of a platform runaway current, and an optimal content and ratio of the deuterium-argon/neon gas mixture are existed for effective runaway current suppression. Within the range of the pre-disruption plasma current (Ip) discussed in this study, the amounts of neon/argon and deuterium in the gas mixture should be 0.50%~0.70% and 1020~1021 m-3, On fusion-reactor-scale tokamak devices with Ip as high as 10 MA, the amount of the injected gas mixture must reach 1022 m-3, which cannot be achieved under the current massive gas injection (MGI) technique.ConclusionsThe pre-disruption plasma current (Ip) is the key factor that influences a runaway current. The larger Ip is, the larger is the runaway current that is formed and more amount of the gas mixture must be injected. On fusion-reactor-scale tokamak devices with Ip as high as 10 MA, the amount of the injected gas mixture must reach 1022 m-3, which cannot be achieved under the current massive gas injection technique. Injecting a deuterium-argon/neon gas mixture through a shattered pellet would be a more viable approach.

BackgroundThe passive cooling system for the reactor cavity of the molten salt reactor (MSR) is an important guarantee to ensure the safe operation of the reactor, is one of the four engineered safety features for the MSR, and its structure design is an important part of the thermal hydraulic design.PurposeThis study aims to find out a suitable passive reactor cavity cooling system (RCCS) to meet the requirements of thermal shielding design for the lower reactor cavity of the MSR, and maximize the removal of reactor core decay heat under accident conditions.MethodsFirstly, based on the design parameters of a 153 MWt MSR, a 1/4 geometric model of the lower reactor cabin of this MSR was established. Then, ANSYS FLUENT 20.1 software was employed to conduct three-dimensional numerical simulation of the flow filed and temperature filed for the lower reactor cabin, the influence of thermal shielding of the lower cavity was analyzed by changing the structure and layout of the passive RCCS, the structure sizes of the passive RCCS with double channel, the thickness of thermal insulation cotton on the intermediate thermal shielding plate and the position of the air inlet pipe. Finally, a new and suitable structure of passive RCCS was proposed after step-by-step improvements for a 153 MWt MSR.ResultsThe simulation results show that the optimized new-style passive air-cooling system with a double channel in the lower reactor cavity is the best among the three structures. Changing the width of the RCCS has little effect on the thermal shielding results of the lower reactor cabin whilst increasing the thickness of thermal insulation cotton on the intermediate thermal shielding plate of the RCCS can significantly reduce the temperature of the inner surface of the concrete wall. The closer the inlet position of the air inlet pipe is to the top of the RCCS, the better the thermal shielding effect. Based on above results, a new-style passive air-cooling system with double channel in the lower cavity is designed to completely dsatisfy the requirements for shielding cooling for the lower reactor cavity of a 153 MWt MSR.ConclusionsThe results of this study provide an important reference for the further engineering optimization design of passive residual heat removal system in the hundred megawatt-scale molten salt reactor.

BackgroundThe extraction of 99Mo medical isotopes from off-gas of molten salt reactors (MSRs) has attracted significant attention. However, previous studies focused only on the estimation of 99Mo nuclide production in the reactor core, and limited studies have focused on the expansion of the entire reactor system.PurposeThis study aims to estimate 99Mo production in the off-gas of a molten salt reactor.MethodsIn this study, a 99Mo migration model with noble metal deposition and a flow effect in an MSR was established by using scientific computing software “Mathematica”, and a corresponding verification based on the deposition result of a molten salt reactor experiment (MSRE) was implemented. Subsequently, an analysis was conducted on the impact of changes in the yield, specific activity, and operational status of 99Mo crude product in MSRE off-gas on the stability of production.ResultsThe analysis results of Mo transport in the MSRE show that the production of 99Mo in the off-gas can reach 69.30 TBq with a specific activity of 0.39 PBq?g-1 under normal operating conditions. In addition, an increase in the removal rate is conducive to the effect on the specific activity of 99Mo whilst the working conditions of bubble surface ratio bubble ratio and oxidation-reduction potential changes have little effect on the specific activity of the product.ConclusionsThe 99Mo in off-gas has high yield and specific activity, and the effects of the conventional operating conditions on the specific activity of 99Mo are beneficial, which indicates that this method can be a potential alternative for 99Mo production.

BackgroundMicro-reactors can be used as a lunar surface power or spacecraft power source for space exploration. Before launching the reactor, a safety analysis should be conducted to prevent a launch accident. Currently, the safety analysis of the radioactive isotope power system does not fully include the safety analysis of the reactor. The main critical safety analysis scenario is that the reactor falls and hits the concrete ground from a high altitude. The reactor may return to criticality after high-speed impact.PurposeThis study aims to investigate the nuclear safety characteristics of a space reactor subjected to dynamic shock under high-speed impact conditions.MethodsFirst of all, based on internal and surface unstructured grids, two simplified reactor models corresponding to two high-speed impact scenarios, i.e., pure fuel reactor vertical impact with ground, and cylinder reactor with a reflector layer and shielding layer impact the ground at a 30° angle were established. Then, the ABAQUS finite element method and unstructured mesh Monte Carlo method of particle transport were combined to predict the criticality properties of the pure fuel and cylindrical reactor during high-speed impact. Based on the surface and internal unstructured mesh Monte Carlo transport technology, the criticality safety analysis platform of micro-reactor under high speed impact was established.ResultsThe results show that the keff induced by the deformation may increase with time for the above mentioned two simplified reactors. The maximum increase in the keff of the pure fuel reactor can reach 1 000×10-5, whereas for the cylinder reactor, the keff is improved to a maximum of 200×10-5. Considering the non-uniform density effect, reactivities of -666×10-5 and -132×10-5 are introduced into the two reactors.ConclusionsThe critical safety characteristics of the reactor under different impact conditions should be evaluated to ensure sufficient safety margins under such accident conditions.

BackgroundMolten salt reactor is one of the six internationally recognized and recommended fourth generation reactors that can use liquid nuclear fuel. The production, transportation, and storage of nuclear fuel involve different processes from conventional solid-state nuclear fuel reactors.PurposeThis study aims to perform criticality safety analysis for nuclear fuel storage of molten salt reactor in compliant with requirements of nuclear fuel management and nuclear safety supervision.MethodsThe design parameters of MSRE (Molten Salt Reactor Experiment) were referred for criticality safety analysis, and the impact of different factors on nuclear fuel salt storage was analyzed. Calculation was completed by selecting storage modeling, critical parameter analysis, and Monte Carlo neutron transport software simulation for liquid fuel molten salt reactor nuclear fuel. The keff values of dry environment storage and water flooded environment storage under the design model were summarized, so did the changes in the total mass of fuel salt.Results & ConclusionsThe quality of subcritical safety control under different conditions are obtained and compared with corresponding raw material salts, intermediate products, and consideration of container walls. This study combines legal regulations and the process of nuclear material circulation for analysis and discussion, summarizes the critical safety characteristics of nuclear fuel salt, and for the first time proposes relevant supervision and evaluation points from the perspective of nuclear safety supervision.

BackgroundNatural circulation systems are widely used in the nuclear industry for decay heat removal, post-accident containment cooling, and the cooling of radioactive waste storage facilities. Because a large number of spent fuel assemblies are densely arranged in the spent fuel pool, the spent fuel assemblies in the central area of the spent fuel pool can be approximated as having adiabatic boundary conditions. In the event of a coolant loss accident, the heat generated by the spent fuel assemblies can be exported only by the natural circulation of air.PurposeThis study aims to develop analytical models for fuel assembly natural circulation characteristics independent of the results of computational fluid dynamics (CFD) to investigate the natural circulation characteristics of spent fuel pools following a coolant loss accident.MethodsBased on the experimental results of a fuel assembly pressure drop of a full-scale pressurized water reactor (PWR), the Darcy–Forchheimer model was firstly revised to predict the pressure drop of humid air flowing through the fuel assembly. Models for the fuel assembly natural circulation flow rate and peak cladding temperature were then established, and the effects of the relative humidity of air on the models were considered. Finally, these models were applied to investigating effects of total heating power, ambient temperature, and relative humidity on the natural circulation flow rate and peak cladding temperature of the fuel assembly.ResultsThe results show that the models accurately predict the natural circulation flow rate and peak cladding temperature of the fuel assembly under different heating powers, and the errors are less than 25% and 20%, respectively, as compared with the experimental measurement values.ConclusionsThe developed models in this study can be used to study the natural circulation characteristics of full-scale PWR fuel assemblies.

BackgroundPencil beam scanning (PBS) proton systems with a rotating gantry makes proton intensity modulation technology widely used for better protection of organs at risk, but it also brings greater challenges to the quality control of proton system.PurposeThis study aims to verify the accuracy of an XRV-124 cone-shaped scintillation detector and to investigate its feasibility for use in PBS quality assurance (QA) procedures.MethodsFirstly, according to the measurement principle of XRV-124, a single setup was implemented to measure the spot size, beam–X-ray coincidence, gantry angle, star-shot test, and mechanical accuracy simultaneously. Then, the spot size and beam–X-ray coincidence were determined at the isocenter using the XRV-124 for 70~240 MeV at increments of 10 MeV and at 30° increments for all gantry angles. The coincidence of the X-ray system and proton beam was evaluated in the X and Y directions of the International Electrotechnical Commission coordinate system (IEC). Finally, the spot size and coincidence results were compared with those of the widely used Lynx detector, whereas the gantry angle results were compared with those of the proton treatment console (PTC).ResultsThe spot sizes obtained using the XRV-124 and Lynx are in good agreement within 0.10 mm for each energy, and the results show the same trend with a maximum deviation of 0.18 mm and 0.11 mm in the IEC-X and -Y directions, respectively. The gantry angles are less than 0.2° compared to those of the PTC. For the star-shot test, the average 3D and 2D distances from the isocenter are 0.4 mm and 0.2 mm, respectively, meets the quality control requirements. The QA items can be completed in 90 min.ConclusionsThis method of this study has been successfully applied to the QA of Varian ProBeam proton radiotherapy system in Hefei ion medical center, indicating that XRV-124 cone-shaped scintillation detector can be widely used in the same type of proton radiotherapy system to improve the QA efficiency and reduce the human error caused by frequent positioning.

BackgroundInterfacial area concentration (IAC) is a key parameter of the interface transfer term in the closed two-fluid model of two-phase flow, which characterizes the strength of the gas-liquid interface transport capacity. There are usually some methods for modeling and predicting the interface area concentration, such as empirical correlation formula and interface area transport equation, but these methods have large data dependence.PurposeThis study aims to provide direction for model revision and improve the prediction accuracy of IAC by adding interpretability to the neural network model.MethodsThe prediction model of IAC based on a neural network was firstly established for better prediction of IAC with two-phase flow. Then, different bubble behavior, physical relationships, and statistical distribution were combined, and the predictive ability of the neural network model with different input feature combinations was compared and analyzed by the post-interpretability method. Finally, based on the structure parameter size of each layer of the neural network, the appropriate data preprocessing method was selected by analyzing the output proportion.ResultsThe post explanatory analysis show that the maximum prediction accuracy of the neural network reaches 95.62% when the inputs of the neural network are the gas superficial velocity (jg), liquid superficial velocity (jf), and void fraction (α). The void fraction is an important factor in IAC prediction, and logarithmic transformation preprocessing of training data can significantly improve the model's predictive ability for real data.ConclusionsThe results of this study provide reference for future interpretability research on interface area concentration.