BackgroundThe electrostatic lens plays an important role in obtaining high quality focused electron/ion/slow positron beams with high spatial resolution and brightness. A novel electric lens, composed of simplified electrode structures with tube diameter gradually decreasing, is proposed for focusing charged particle beams with low energy and large spot size.PurposeThis study aims to investigate the beam dynamics of this designed electrostatic lens for validation.MethodsBased on overall structure of electric lens focusing system, the charged particle optical simulation software SIMION was employed to optimize the parameters of this focusing system. Then combined with electron beam experiments, the key technologies involved in the electronic lens were studied in detail, including influencing factors, their distributions, and the focusing performance of the lens.ResultsThe results show that the large transverse space of the initial electron beam can be compressed effectively by arranging the electrode structure and electric potential of the lens, with focusing efficiencies exceeding 80%.ConclusionsThe focusing method proposed in this study has significant lateral compression advantages with wide application prospects in many focusing scenarios of different charged particle beams, such as reactor positron sources.

BackgroundTitanium and its alloys are widely utilized within the military, aerospace, shipping, nuclear energy, and biomedical fields because of the advantages of low density, high strength, good corrosion resistance, and high biocompatibility. Moreover, titanium films are important materials for surface protection due to their high hardness and good compactness. During the preparation of titanium films, the surface morphology and phase structure will be influenced by substrate properties (e.g., surface morphology and temperature), working gas pressure, and other factors. Substrate temperature mainly influences the growth process of thin films, which directly affects the grain structure of the films, and thus changes the corresponding mechanical properties.PurposeThis study aims to establish the relationship between substrate temperature and mechanical properties of titanium films.MethodsFirstly, titanium film samples were prepared at a substrate temperature range of 600~750 ℃ by using resistance evaporation coating on the surface of molybdenum substrate. Then, the structural characterization of the film was examined using X-ray diffraction (XRD) so as to obtain the preferred orientation of titanium film. Scanning electron microscopy (SEM) and atomic force microscopy (AFM) were employed to identify the surface morphologies of the titanium films, including grain size distribution and surface roughness. Finally, an AFM nano-indentation method was performed to examine the mechanical properties of the titanium films and obtain the elastic modulus of titanium film.ResultsThe results demonstrate that the substrate temperature significantly influences the microstructure and mechanical properties of titanium films. When the substrate temperature increases from 600 ℃ to 750 ℃, the preferred orientation of titanium films changes from (101) to (002) due to the competition between the minimization of strain energy and surface energy. With the increase of substrate temperature, the mobility of titanium atoms on the substrate increases, resulting in increased average grain size, surface roughness, and elastic modulus of the titanium films. The average grain size increases by 26% as the substrate temperature increased from 600 ℃ to 750 ℃.ConclusionsThe microstructure, surface morphology, and mechanical properties of titanium films are sensitive to substrate temperature. A high substrate temperature in the process of resistance evaporation is more desirable to obtain titanium films with high mechanical properties.

BackgroundIn the realm of cosmological ray studies employing plastic scintillation fibers, it is essential to conduct quantitative analyses of the photon number from the fiber's output pulse for the successful design of readout electronics.PurposeGiven the absence of a weak cursor setting device such as a single photon source. This study aims to quantitatively analyze the number of photons generated by photon incidents within the fiber calibrating without a weak photon source such as a single photon source.MethodsFirstly, photon numbers within weak optical pulses induced by muons in optical fibers with diameters of 1 mm and 2 mm were determinated by the calibration method that making use of inherent non-photogenerated carrier characteristics of the silicon photomultiplier tube (SiPM). Then, the Geant4 software was employed to simulate the theoretical photon yield of muons in these optical fibers, and the simulation results were compared with experimental data for validation.ResultsThe verification results indicate that the anticipated photon count in the optical pulses within fibers with diameters of the 1 mm and 2 mm fibers are 44 and 85, respectively. The deviation from the simulation results is 4.55% and 10.59%, respectively.ConclusionsThe results validate the efficiency of the low photon number measurement method, demonstrating its ability to accurately measure the photon count generated by the incident fiber without the need for additional calibration equipment. This method may extend to other scenarios that require the measurement of photon numbers in weak light pulse situations.

BackgroundCosine function shaping (or cos shaping) is used to digitally shape nuclear pulse signals, as the shaping method is simple and has high operability and flexibility.PurposeThis study aims to explore different cosine shaping methods of nuclear pulse signals, and evaluate their effect.MethodsFirstly, based on the single exponential decay signal and cosine pulse signal, transfer functions and cascade formulas of three different cosine shaping methods in the Z-domain were derived. The influence of the parameter selection on the shaping effect in the cosine shaping algorithm was analyzed. Then, the cosine shaping methods were developed for the simulated nuclear signals and the actual sampled nuclear signals, and the cosine shaping, amplitude extraction, and energy spectrum construction of the digital nuclear signals were implemented in the field programmable gate array (FPGA) system. Finally, the gamma (γ) energy spectrum of 137Cs (NaI(Tl) detector) was evaluated using the different cosine shaping methods.ResultsThe results of γ energy spectrum from 137Cs (NaI(Tl) detector) demonstrate the satisfactory performance of all three digital cosine shaping algorithms in terms of energy resolution and counting. The symmetric zero-area cosine shaping performance index is improved relative to conventional methods.ConclusionsThe three kinds of digital cosine shaping methods all achieve accurate cosine shaping for simulated and real nuclear signals. The three cosine shaping methods proposed in this study may be applied to shape functions in other research areas.

BackgroundThere are no commercially available channel electron multipliers (CEMs) made of glass in domestic market of China; more complex CEMs with helix channels are scarcer.PurposeThis study aims to develop a CEM with a single helix channel, and test its performace for satisfying the requirements of high-end users of such products.MethodsFirst of all, a series of manufacturing process designs and improvements were made on the basis of the formula of microchannel sheet glass, resulting in the production of a single spiral channel electron multiplier with suitable performance. Then, a CEM analog mode test device with a disc-incense type tantalum filament as the input current and a CEM pulse-counting mode test device with an ultraviolet light-emitting diode combined with a gold cathode as the input signal were set up to conduct comprehensive testing of the device's performance parameters.ResultsThe newly developed CEM with single helix channel achieves gains of 1×104~1×106 in the analog mode and 1×107~1×108 in the pulse-counting mode. The gain value increases with the increase of the working voltage, and the rise time of the output pulse is 2~3 ns.ConclusionsThe overall performance of the developed CEM is close to that of foreign counterparts, and the CEM can be used in related instruments.

BackgroundDirect-current radiofrequency (DC-RF) hybrid plasma has broad application prospects in the field of nuclear ultrafine powder material preparation owing to its characteristics of high temperature and high chemical activity.PurposeThis study aims to explore the flow and heat-transfer characteristics of DC-RF hybrid plasma, so as to provide references for the design and stable operation of the plasma generator device.MethodsFirst of all, the hybrid plasma generator was assumed to be a two-dimensional axisymmetric model, and the device was filled with pure argon plasma in a local thermodynamic equilibrium (LTE), steady, and turbulent flow state. Then, the ANSYS FLUENT software was employed to establish a two-dimensional model for the DC-RF hybrid plasma torch structure, and the spatial distributions of the temperature and flow field in DC-RF hybrid plasma torch were simulated using the k-ε turbulence model with SIMPLE algorithm based on velocity and pressure coupling solver. Finally, the effects of changes in the operating parameters were analyzed based on these results.Results & ConclusionsThe simulation results indicate that increases in the DC arc current, reaction gas flow rate, and cooling gas flow rate can reduce backflow effects at the entrance of hybrid plasma torch. The temperature and area of the plasma arc near the RF coil increase with the RF coil current. However, an excessive current and gas flow rate may adversely affect the operation of the device. Various requirements of material handling processes on the premise of stable operation of the device can be satisfied by adjusting working parameters for the control of the hybrid plasma flow field profiles.

BackgroundThe use of controlled X-ray sources instead of 137Cs radioactive sources in density logging has become a new trend. The high voltage on the target substrate significantly affects the intensity of the X-ray source, 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 examine the depth-of-investigation characteristics and influence of a 350-kV high-voltage X-ray density logging instrument.MethodsThe depth of investigation of various source distance detectors in 20% water-bearing limestone formation was studied using the Monte Carlo method. By comparing the investigation characteristics of 350-kV high-voltage X-ray source and 137Cs source density logging, the reasons for the differences in the depth of investigation among them were analyzed. Moreover, the contribution of mudcake and formation to the detector and the density deviations of various detectors were analyzed via simulation. Finally, the influence of mudcake on the density logging response of the well wall was explored.ResultsThe results indicate that the depth of investigation of X-ray density logging instrument increases with the augment of source distance. Compared to the 137Cs source density logging, the scattered particles of the X-ray density logging are mainly concentrated at 1~3 cm from the bore wall, resulting in the depth-of-investigation differences between the two techniques. Furthermore, the contributions of mudcake and formation to different source distance detectors are different, and the detector density deviation decreases with the increase in source distance.ConclusionsThis study affords a theoretical basis for the depth-of-investigation characteristics and influence of 350-kV high-voltage X-ray density logging.

BackgroundThe high-temperature liquid lead-bismuth metal has a scouring and wear effect on the head of the axial flow lead-bismuth pump impeller blades in a lead-cooled fast reactor system, causing the protective layer on the blade surface to break down and material corrosion rate to accelerate.PurposeThis study aims to reduce the scouring wear effect of the high temperature liquid metals on blade surfaces.MethodsFirst of all, three types of leaf top clearance structures, i.e., plane, chamfered right angle, and chamfered rounded angle, were designed. Then, the reliability of the simulation results was verified by the scaling conversion method, and the commercial CFD software ANSYS CFX with SST k-ω turbulence model was employed to analyze the variation of flow velocity, shear force, and flow pattern with scouring and wear characteristics under different leaf top clearance structures. Finally, the energy loss of the high temperature liquid lead-bismuth metal on the material surface was analyzed using the wall entropy yield.ResultsAnalysis results show that the head and efficiency of the chamfered right angle model are reduced by 1.02% and 0.64%, respectively, compared with those of the flat surface under standard operating conditions, and the chamfered angle model shows a 0.51% reduction in the head and 0.51% efficiency increase. The impeller scouring wear effect occurs predominantly near the inlet edge of the blade rim, and the effect of the high temperature liquid metal on the blade head scouring wear is improved by the chamfered and rounded designs.ConclusionsThe chamfered and rounded designs reduces the mechanical energy loss on the blade surface by reducing the flow velocity at the top clearance and reducing the scouring wear effect at this location. Therefore, the rounded design and the right angle design could improve the influence of high-temperature liquid lead-bismuth metal on the erosion wear of the blade head.

BackgroundThe experimental advanced superconducting tokamak (EAST) feeder system is an essential part of the device that connects the superconducting magnet and high-temperature superconducting current lead. It provides the magnet with feeding and energy release channels in the event of quenching. In recent years, owing to the increase in the inlet temperature of the toroidal field (TF) feeder system, the outlet temperature has occasionally exceeded the threshold of 6.1 K, resulting in the termination of the experiment.PurposeThis study aims to ensure the continuation of the EAST experiment by thermal stability analysis under new TF feeder outlet temperature threshold of 6.5 K.MethodsBased on the system structure and low-temperature operation data of the TF feeder and superconducting conductor, a mathematical model for temperature margin and current shunt temperature of superconducting conductor was established. Then, the mathematical model and GANDALF software were employed to calculate the temperature and stability margins of superconducting conductors during operation under different background magnetic fields and operating currents.ResultsThe calculations results indicate that under the new threshold of 6.5 K, the temperature margin of the conductor is greater than 1.5 K and that the stability energy margin is greater than 200mJ·cm-3.ConclusionsThe superconducting conductor remains safe for use after increasing the threshold.

BackgroundThe primary coolant flow rate is essential in preventing departure from nucleate boiling. The implementation of a low-leakage core loading pattern in advanced passive (AP) technology-based nuclear power units has increased the temperature difference gradient at the core outlet, resulting in elevated uncertainty in the flow rate calculations when using the heat balance method.PurposeThis study aims to validate a measurement and calculation method based on the Bernoulli equation model for accurately determining the primary coolant flow rate in AP nuclear power units, hence meeting the design and regulatory requirements.MethodsFirst of all, measurements were conducted for the primary loop main equipment and bend pipe flowmeter pressure differentials during the commissioning phases. Calorimertic balance tests were performed at power levels of 50%, 75%, 90%, and 100%. Then, the bend pipe flowmeter coefficients were calibrated using the flow rate values obtained from the hot function test and 100% rated thermal power (RTP). Finally, based on weighted factors, the total flow rate values for the reactor coolant system (RCS) were calculated with emphasis on the minimization of uncertainties.ResultsThe proposed measurement and calculation method yields primary coolant flow rate values with a relative error of less than 4%. The total flow rate after loading is within the range of 95.8% to 104% of the expected optimum flow rate. The uncertainty of the volumetric flow rate calculated from NAPs is lower than 1.9%, demonstrating a novel approach for precise measurements in other units.ConclusionsThe method of this study offers an advanced perspective for reactor coolant precise measurements in other units, with primary coolant flow rate values exhibiting minimal relative error and volumetric flow rate values from NAPs demonstrating low uncertainty.

BackgroundThe recoil release caused by the collision of a neutron with a target nucleus has a significant impact on the analysis of activated corrosion product sources in a reactor. In water cooled reactors, a recoil release in the irradiated area can cause the activation corrosion products to leave the wall surface and enter the coolant, which then migrates to the non-irradiated area with the flow of coolant, thereby making the non-irradiated equipment also radioactive.PurposeThis study aims to analyze the influence of recoil release on the source term of activated corrosion product in reactor.MethodsBased on the investigation of the mode of action of recoil release in a reactor, a calculation model and a program module for recoil release was established and integrated into the CATE program. Then, the effects of recoil release on the concentrations of activation corrosion products in the nuclear reactor were analyzed by using two approaches, one involved specifying a recoil release probability, while the other involved dynamically calculating the recoil release probability. Finally, values of the main activated corrosion products nuclides 58Co and 60Co in the core and steam generator before and after considering recoil release were calculated, and the impact of recoil release on the activation corrosion products and its implications for the actual operation of the reactor were explored.ResultsThe calculation results indicate that the recoil release probability decreases from 45% at the beginning of the simulation to 0.3% at the end of the simulation. However, the variation pattern of the activity ratio of 58Co and 60Co in the core and steam generator remains the same as that without recoil release. The activity ratio is 91% and 203% respectively, compared to the case without recoil release.ConclusionsThe total probability of recoil release is related to the thickness of the corrosion product layer and gradually decreases with the operating time of the reactor.

BackgroundThe magnitude of the thermal stress in the first wall system is one of the key factors affecting the safe operation of the fusion reactor.PurposeThis study aims to investigate the effect of a rough substrate on the thermal stress in the W/316L stainless steel first wall system.MethodThe finite-element analysis software Ansys Workbench was employ to analyze the distribution of thermal stree in a W/316L stainless steel first wall system with a rough substrate. Depth analysis was conducted on factors such as the temperature, coating thickness, and substrate thickness that affect the magnitude of thermal stress. Meanwhile, starting from the interface shear stress in the system, the influence of rough substrate on the bonding strength of coatings was simultaneously investigated.ResultsThe simulation results indicate that the thermal stress in the rough substrate system increases with the the increase of temperature and substrate thickness, but decreases with the increase of coating thickness. The maximum thermal stress and the adhesive strength between the coating and the substrate are raised by the introduction of the rough substrate.ConclusionsResults of this study can provide reference for the development of high-adhesive strength first wall coating systems.

BackgroundInert matrix fuel (IMF) can efficiently convert plutonium and long-lived minor actinides used for preventing the proliferation of nuclear weapons and improving spent fuel disposal, hence has been becoming a hot research topic in recent years. The sol-gel method has the advantage of uniform elemental distribution of the products and the wet operation process is less likely to produce radioactive dust, therefore, it has been used to prepare zirconium-based IMF in the research.PurposeThis study aims to prepare a colloidal solution with good dispersive properties and to obtain IMF microspheres with good sphericity, uniform size, and homogeneous elemental distribution.MethodsFirst of all, ThxZr1-xO2 inert matrix fuel was prepared by an external gelation process, and the sol-gel viscosity was used as the main gelation index. Then, the variation tendency of sol viscosity with c(NH4+)/c(NO3-) was investigated for different metal ions concentrations and different temperatures. Finally, the statistical distributions of colloidal particle sizes were obtained for different metal ions and reaction temperatures by laser particle sizing tests, and the X-ray diffraction (XRD) was used to study the structure of IMF after heat treatment at different temperatures.ResultsThe results showed that the complex gelation parameters and properties can be categorized and quantified using gelation field diagrams. In addition, ThxZr1-xO2 IMF kernels with uniform element distribution, good sphericity, and integral appearance were obtained by optimizing the process parameters. Zirconia showed low solubility behavior in the thorium-oxide system, leading to the generation of a biphasic structure.ConclusionsThe results of this study indicate that zirconium-based spherical IMF microspheres with good performance can be prepared by external gelation method.

BackgroundDouble-layered heat exchanger tubes can effectively reduce the probability of heat exchanger tube rupture accidents from design; however, the thermal contact resistance between tubes may decrease the heat transfer efficiency of the heat exchanger tubes. This is not conducive to the smooth heat export of the first circuit system of lead-bismuth reactor; therefore, there is an urgent need to optimize the design of heat exchanger tubes.PurposeThis study aims to optimize the structure of the double-layered tubular main heat exchanger, so as to improve its heat transfer efficiency.MethodsFirstly, the double layer heat exchange tube type main heat exchanger of lead bismuth reactor was taken as the research object, the interstices of double-layered heat exchanger tubes was filled with gallium-based graphene nanofluid as the thermal interface material. Then, the influence of heat exchanger tube length, wall thickness, outer diameter, and spacing on the heat transfer performance were analyzed and results were compared with that of a double-layered tubular main heat exchanger without the thermal interface material. Finally, taking the JF factor and cost efficiency ratio (CER) as optimization objectives and using the aforementioned four parameters as optimization variables, the heat transfer performance of the main heat exchanger was optimized and evaluated on the basis of a genetic algorithm, hence to obtain a new double-layered heat exchanger design for a lead-bismuth reactor.ResultsThe results of comparative calculations indicate that the total heat transfer coefficient, first-loop pressure drop, JF factor, and CER factor of optimized main heat exchanger increase by 5.79%, 2.32%, 5%, and 24.62%, respectively.ConclusionsFilling the gap of double-layered heat exchanger tubes with a gallium-based graphene nanofluid can effectively improve the heat transfer performance of the double-layered tube exchanger while reducing the accident probability of steam generator tube rupture.

BackgroundLoss of Coolant Accidents (LOCAs) is a crucial research topic for nuclear reactor safety analysis, and understanding the thermal–hydraulic behavior of the rod bundle channels during the reflooding stage of a LOCA is essential.PurposeThis study aims to develop theoretical models of the reflooding stage in addition to providing benchmark data for evaluating the safety analysis code for LOCAs in a reactor and for the design of the residual heat removal system.MethodsA series of bottom reflooding tests were conducted on a 5×5 rod bundle in the film boiling test facility at the nuclear safety and operation laboratory (NUSOL) of Xi'an Jiaotong University using uniformly heated rods. The experimental results were analyzed in detail, and the surface parameters of the heated rod bundle were obtained by solving a one-dimensional transient inverse heat conduction problem. The effects of different experimental conditions on the velocity of the quench front propagation were investigated. Furthermore, the experimental results were compared and calculated using the thermal safety analysis code, and the problems with the thermal safety analysis code RELAP5 reflooding simulation are summarized.ResultsOur results indicate that a high inlet flow rate, high inlet subcooling degree, and low power density are favorable for the propagation of the cold front during the reflooding process. Additionally, the root mean square (RMS) error of the simulated quench time and peak cladding temperature (PCT) are 40.994 s and 61.465 K, respectively. However, the simulation results have a relatively large error compared with the experimental results in the post-critical heat flux (CHF) heat transfer stage, primarily owing to the issues with the boiling mode judgment and membrane boiling heat transfer model.ConclusionsThe experimental data of this study can serve as new verification data for flow and heat transfer prediction models during the reflooding process; it can also be used to evaluate and optimize the thermal-hydraulic safety analysis code. Loss of Coolant Accidents (LOCAs) is a crucial research topic for nuclear reactor safety analysis, and understanding the thermal-hydraulic behavior of the rod bundle channels during the reflooding stage of a LOCA is essential.