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Methods study on high-resolution bunch charge measurement based on cavity monitor
Shanshan CAO,Yongbin LENG,Renxian YUAN,Longwei LAI,Jian CHEN
Nuclear Techniques. 2021, 44 (4):
40101-040101.
DOI: 10.11889/j.0253-3219.2021.hjs.44.040101
BackgroundBunch charge is the most fundamental characteristic parameter of the particle beam. The Shanghai Soft X-ray Free Electron Laser Device Facility (SXFEL) needs to add a charge feedback loop to accurately control the bunch charge and the resolution of a real-time single bunch charge measurement is required to be better than 0.5%. At present, commonly used monitors that can realize online non-intercepting bunch charge measurement include beam current transformers, button-type or strip-type electrodes, and cavity-type probes, etc. Among them, the beam current transformer is susceptible to various electromagnetic interference, and its resolution is difficult to be improved; the button probe and the stripline probe have obvious bunch position dependence. The cavity probe has high sensitivity and high signal-to-noise ratio (SNR), position independence under paraxial conditions, thus it is very suitable for high-resolution bunch charge measurement. PurposeThis study aims to explore the methods of high-resolution bunch charge measurement based on cavity monitors. MethodsBased on the traditional cavity probe signal measurement system using an external local oscillator signal mixing scheme, a new dual-cavity mixing scheme with a simple system structure without local oscillator frequency synthesis was proposed. It could work independently without synchronization timing signals. Both schemes were evaluated by beam experiments in the SXFEL facility. ResultsThe experimental results show that the resolution of the single-cavity external IF (intermediate frequency) mixing scheme is better than 0.07% while the resolution of the dual-cavity mixing scheme is better than 0.2%. ConclusionsBoth schemes can achieve a high-resolution bunch charge measurement and satisfy the requirements of SXFEL user facility.
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Simulation study on the performance of micro X-ray tube with diamond optical window
Yiqiang XING,Jiankun ZHAO,Weicheng LI,Yibao LIU,Wei LIU,Shuang JIANG
Nuclear Techniques. 2021, 44 (4):
40201-040201.
DOI: 10.11889/j.0253-3219.2021.hjs.44.040201
BackgroundAs a hot material, diamond has been widely used in various civil and military equipment with high hardness (Mohs hardness is 10), compressive strength (greater than 1.2 GPa), excellent thermal (thermal conductivity at room temperature is 20~22 W·cm-1·K-1, thermal expansion coefficient at room temperature is only (1.1~1.3)×10-6 K-1), optical (higher transparency to X-rays) and electrical properties. Traditional optical window material beryllium is harmful to the human body and industrial hazards, its hardness, thermal conductivity and the thermal expansion coefficient are much lower than that of diamond. PurposeThis study aims to explore the performance of micro X-ray tube with diamond optical window by simulation, obtain the best thickness of diamond for X-ray tube optical window and its shielding effect against low, medium and high energy X-ray. MethodsThe Monte Carlo method was used to calculate the effective transmission ratio, peak back ratio and transmission ratio of characteristic X rays with diamond optical window in high energy region. The X-ray tube with 50 kV tube voltage and 1.0 mA tube current was taken as an example to determine the optimal thickness of diamond. ResultsSimulation results show that the Kα characteristic X-ray effective transmittance and peak to total X-ray ratio of the silver target increase continuously with the increase of the diamond window thickness, and the optimum diamond thickness is 2.0 mm. Under the optimal thickness of 2.0 mm, the effective transmittance is 154.5%, high-energy X-ray transmittance is 74.5%, and the characteristic X-ray peak back ratio is 27.9%. ConclusionsDiamond can be used as a substitute for beryllium optical windows and has good application prospect.
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Study of beam model of medical linear accelerator based on virtual single point source
Huijuan LI,Shengxiu JIAO,Zhongben CHEN,Xiaowei LIU
Nuclear Techniques. 2021, 44 (4):
40202-040202.
DOI: 10.11889/j.0253-3219.2021.hjs.44.040202
BackgroundThe beam model of medical linac is one of the bases of dose calculation. It is of great significance to study the beam model of medical linear accelerator in clinical application of radiation dose calculation. PurposeThis study aims to provide a virtual point source model of medical linear accelerator X-ray beam, and to discuss the applicability of the model by fitting the measured dose distribution. MethodsThe X-ray beam was divided into primary ray and scattered ray. The primary ray was modeled as an anisotropic point source whilst the scattered ray was modeled as a point source whose intensity was related to the radiation field. The model parameters were obtained by fitting the measured absorbed dose distribution in water phantom. ResultsFor Varian Trilogy 6 MV linear accelerator, the dose distribution given by the fitting model is in good agreement with the measured dose distribution, and the difference between the measured value and the fitting value of the total scattering factor (SCP) is less than 1.5%. For 3 cm × 3 cm, 10 cm× 10 cm and 40 cm× 40 cm radiation fields, from 1.4 cm (maximum dose depth) to 20 cm underwater depth, the average differences between the fitting and the measured value are 1.3%, 1.3% and 1.1% respectively. The average differences between the fitting and measured values of off-axis dose in the area with off axis dose ratio (OAR) greater than 90% are 2.35%, 0.75% and 0.53% respectively at the depth of 10 cm. ConclusionsThe virtual point source model of X-ray beam can well reconstruct the absorbed dose in phantom.
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Design of real-time feedback system of dynamic current for spot scanning in medical heavy-ion facility
Haoyu JIANG,Jiang ZHAO,Zhongzu ZHOU,Dezhi WANG,Daqing GAO
Nuclear Techniques. 2021, 44 (4):
40203-040203.
DOI: 10.11889/j.0253-3219.2021.hjs.44.040203
BackgroundSpot scanning therapy is an accurate radiotherapy method in medical heavy-ion facilities. The process of spot scanning therapy is that the current of scanning power supply changes rapidly in real-time. The accuracy control of beam position depends on the repeatability of the scanning power supplies and the stability of the model parameters during spot scanning therapy. However, the amplitude of position current is small and changes rapidly, it is easy to be submerged by noise, which makes it difficult for the treatment terminal to obtain accurate scanning dynamic current remotely. PurposeThis study aims to design and implement a real time feedback system of point scan position current for quick analysis and optimization of the accuracy and repetition of the point scanning position current for medical heavy-ion facility. MethodsThe Xilinx's new scalable processing platform Zynq-7000 SoC (System on Chip) was employed to implement this system by combining high speed Fiber-optic communication with PCIE (Peripheral Component Interface Express) serial bus technology. Both the waveform data acquisition of Direct Coupled Current Transformer (DCCT) and feedback calculation of the remote power supply output current were achieved in the platform. The position current feedback system was tested in field of Lanzhou heavy ion therapy facility. ResultsThe field test results show that the system can obtain the position current data of spot scanning in real time, and the measured position current deviation is 0.106 8 A, satisyfying the design requirement of less than 0.17 A. ConclusionsThe study provides an effective technical means for the precision control and optimization of spot scanning position current.
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Application research of γ energy spectrum analysis based on improved grey wolf algorithm
Wei LIU,Jiankun ZHAO,Yibao LIU,Weicheng LI,Yiqiang XING
Nuclear Techniques. 2021, 44 (4):
40204-040204.
DOI: 10.11889/j.0253-3219.2021.hjs.44.040204
BackgroundTraditional γ energy spectrum analysis methods have the disadvantages of long time, many steps and low accuracy. PurposeThis study aims to propose an improved gray wolf algorithm based on qualitative analysis to obtain nuclide information faster and more accurately. MethodsThe convergence factor based on the inverted S-curve was adopted in the improved gray wolf algorithm to maintain the balance between the global search and the local search of the algorithm, and solve the problem of local optimum that the original gray wolf algorithm fallen into easily. A dimension-by-dimension update strategy was introduced to improve the accuracy and efficiency of optimization. Finally, this improved gray wolf algorithm was applied to experimental test on the gamma spectra of "Soil standard source" and "Marin cup standard source". ResultsExperimental results show that the improved gray wolf algorithm has a better analysis effect on multi-nuclides mixed gamma spectra. The analysis speed is 59% faster than the original gray wolf algorithm.The analytical error of all nuclides in the soil standard source can be controlled within ±10%. The average absolute error of 8 nuclides is within 4%. Except for 226Ra and 235U in the Marin Cup standard source, the errors of other nuclides are less than 10%. The average absolute error of the 8 nuclides is within 7%. ConclusionsThe improved gray wolf algorithm provides a new method for rapid analysis of nuclides.
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Preparation of ZIF-8 loaded citric acid-coated nano-zerovalent-iron and its adsorption properties for U(VI)
Jiang HE,Fei GAO,Feng ZHANG,Feng FENG,Shilong SHI,Jun LIU
Nuclear Techniques. 2021, 44 (4):
40301-040301.
DOI: 10.11889/j.0253-3219.2021.hjs.44.040301
BackgroundThe separation, removal and recovery of U(VI) from water is of great significance to the sustainable development of nuclear energy. PurposeThis study aims to prepare a composite material ZIF-8 loaded by citric acid-coated nano-zero-valent iron (nZVI@CA/ZIF-8), and investigate its adsorption behavior and mechanism for U(VI). MethodsnZVI@CA/ZIF-8 was chemically synthesized and its adsorption properties for U(VI) were analysised by various characterization methods. The effect of time, initial pH, uranium concentration, temperature and ionic strength on Batch U(VI) adsorption were investigated from the aspects of adsorption kinetics, isotherms and thermodynamics. X-ray diffraction spectra (XRD), scanning electron microscopy - energy dispersive spectroscopy (SEM/EDS), X-ray photoelectron spectroscopy (XPS) and Fourier transform infrared spectra (FTIR) were used to characterize and analyse the adsorption behavior and mechanisms. ResultsThe results show that the equilibrium adsorption capacity of uranium is about 115.0 mg?g-1 when the initial uranium concentration is 50.00 mg?L-1 at pH 4.0 and 25 ℃. The U(VI) adsorption amount of nZVI@CA/ZIF-8 increases with the increase of initial pH, initial U(VI) concentration and temperature, and maintains good U(VI) adsorption performances even at high Na+ ionic concentration (0.5 mol·L-1). The equilibrium adsorption capacity of uranium is about (110.0±5.0) mg·L-1 when the initial uranium concentration is 50.00 mg·L-1 at pH 4.0 and 25 ℃. The adsorption is a spontaneous endothermic monolayer chemical process, which can be deduced by the pseudo-second-order kinetic and Freundlich models. Brunauer Emmett teller (BET) measurement suggests that nZVI@CA/ZIF-8 displays a porous structure and a large specific surface area (1 271 m2?g-1). Results of XRD and SEM-EDS demonstrate that the crystal structure and microscopic morphology of ZIF-8 and nZVI are present in the composite. Test results of XPS and FTIR verify that U(VI) can be reduced to U(IV) by nZVI particles exists the nZVI particles could reduce on the surface of composite in the form of UO2. Moreover, nZVI@CA/ZIF-8 can also adsorb U(VI) from the solution by forming Zn-O-U coordination bonds. ConclusionsThis study can provide technical and theoretical references for the synthesis of metal-organic framework composites and the study of its U(VI) removal from radioactive wastewater.
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Study on spectrum stabilization technique for a portable LaBr3(Ce) gamma spectrometer
Chen CHEN,Huan WU
Nuclear Techniques. 2021, 44 (4):
40401-040401.
DOI: 10.11889/j.0253-3219.2021.hjs.44.040401
BackgroundAs one of the most important issues in the field of gamma spectrum analysis, the main function of spectrum stabilization is to suppress the "spectral line drift" caused by environmental changes. Portable LaBr3(Ce) gamma spectrometer is mainly used to analyze and process the gamma spectra of radionuclides and obtain qualitative and quantitative information about radionuclides. It often works in the field complex environment where the spectrum drift is inevitable, hence must have the function of spectrum stabilization. PurposeThis study aims to improve the performance of spectrum stabilization technique for a portable LaBr3(Ce) gamma spectrometer. MethodsThe intrinsic characteristic peak of LaBr3(Ce) detector and the peak of 40K in the natural background were made of as the reference peaks. Then, the concept of "spectrum similarity" was defined to calculate the similarity between the measured spectrum and the standard background spectrum. Finally, a new spectrum stabilization method using standard reference peak and "spectrum similarity" was proposed to measure and correct the deviation. Experiments were carried out verify the performance of this new method. ResultsExperimental results show that the technology can effectively stabilize spectrum in the temperature range of -30 ℃ to 50 ℃ without calibration of the relationship curve between the reference peak and the temperature, and the peak drift of the reference peak is stable within ±1 channel even if the temperature changes more than 50 ℃?h-1. ConclusionsThis technology effectively improves the application range of portable LaBr3(Ce) spectrometers.
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Development of a super-high rang γ monitor for high temperature and pressure environment
Guangzhi SUN,Shunli QIU,Wei XIAO,Le ZENG,Haifeng LIU,Yu PEI,Mengtuan GE,Yulin ZHOU,Hui CHENG
Nuclear Techniques. 2021, 44 (4):
40402-040402.
DOI: 10.11889/j.0253-3219.2021.hjs.44.040402
BackgroundUnder accident conditions of nuclear power plants, radiation measurement is needed in the high temperature, high humidity and high pressure environment in the containment vessel where the ordinary radiation monitor does not work normally. PurposeThis study aims to design and implement a super-high range γ-dose monitor used in the high temperature and high pressure environment of nuclear power station containment vessel. MethodsThe overall design of the monitor, the structure design of the ionization chamber detector, the calculation of the compressive strength of the detector, and the design of a wide range weak current measurement circuit were carried out based on current-voltage-frequency conversion. The ionization chamber detector located in the containment vessel could bear with the lose of coolant accident (LOCA) environment with high temperature up to 200 ℃ and high pressure of 0.7 MPa. The monitor was calibrated by standard radioactive installation and mega radioactive device. Then the LOCA experiment vessel to was employed to test the high temperature and pressure endurance. ResultsTest resultresult shows that the monitor has good sensitivity linearity within the range of 1 mGy·h-1 to 50 000 Gy·h-1, and can provide stable signal under LOCA environment. ConclusionsThe device is suitable for γ radiation monitor under accident conditions.
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Ion pulse ionization chamber for online measurements of the radon activity concentration
Tianli QIU,Meng LI,Xianglun WEI,Herun YANG,Peng MA,Chengui LU,Limin DUAN,Rongjiang HU,Zhoubo HE,Juncheng LIANG,Ming ZHANG
Nuclear Techniques. 2021, 44 (4):
40403-040403.
DOI: 10.11889/j.0253-3219.2021.hjs.44.040403
BackgroundRadon and its daughters are seriously harmful to human respiratory organs. PurposeThis study aims to accurately measure the concentration of radon in ambient air by ion pulse ionization chamber for online measurements. MethodsBased on the principle that alpha rays emitted by radon and its daughters ionize air molecules, a detection system based on ion pulse ionization chamber was developed to record the information of single radioactive particles and accurately measure the concentration in the air. Double shielding structure to adopted to suppress the noise effectively for the detector, and ionized data was obtained by ion pulse collection and oscilloscope waveform sampling. The radon content in the air was measured in different locations, and the effective signals and counts were obtained. Results & ConclusionsThe measurement results show that the full width at half maximum (FWHM) of the 241Am alpha particle energy resolution is 17% under the detector's working voltage range of -400 to -1?200 V. The radon concentrations in the air measured at two campus sites in Lanzhou city are (22.05±0.86) Bq·m-3 and (20.31±0.84) Bq·m-3, respectively, that conform to the provisions of GB/T 50325?2020 on indoor radon concentration.
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Simulation study on time resolution optimization of silicon drift detector
Nian YU,Yupeng XU,Yanke CAI,Yuxuan ZHU,Xiaofan ZHAO,Can CHEN
Nuclear Techniques. 2021, 44 (4):
40404-040404.
DOI: 10.11889/j.0253-3219.2021.hjs.44.040404
BackgroundThe silicon drift detector (SDD) has been widely used in space exploration in recent years. The SDD has outstanding energy resolution, but its time resolution is not very satisfactory. The charge packet generated by the interaction between the incident photon and the detector takes some time to drift to anode. The drift time depends on the distance between the interaction position and the anode, and usually can't be measured directly. Therefore, the uncertainty of drift time is the main factor that affects the time resolution of the SDD system. PurposeThis study aims to propose a method called pulse amplitude ratio to measure the drift time so as to improve the time resolution of the SDD system. MethodsThe correspondence between the drift time and the rise time of the output signal of the charge sensitive amplifier (CSA) was made use of. The rise time was measured by the pulse amplitude ratio of triangular and trapezoidal shaping, hence the drift time was obtained by measuring the pulse amplitude ratio. Matlab/Simulink module and Xilinx module of system generator were used to build the system for simulation of the readout chain for pulse amplitude ratio measurement. The relationship between pulse amplitude ratio and CSA output signal, and the influence of SDD and CSA electronic noise on pulse amplitude ratio were investigated in details. ResultsSimulation results show that the pulse amplitude ratio of the shaping network only depends on the rise time, and the electronic noise of SDD and CSA can cause the measurement error. ConclusionsThe influence of electronic noise of SDD can be reduced by using optimized parameters of the shaping network, and the accuracy of the corrected arrival time can be improved by about one order of magnitude.
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Numerical simulation of controllable source neutron porosity logging based on CLYC detector
Yue ZHOU,Huawei YU,Meng WANG,Benqiang DU,Zhijie LIU
Nuclear Techniques. 2021, 44 (4):
40501-040501.
DOI: 10.11889/j.0253-3219.2021.hjs.44.040501
BackgroundUsing Cs2LiYCl6:Ce (CLYC) instead of 3He detector in neutron porosity logging has become a new idea for the development of logging tools. PurposeThis study aims to investigate the feasibility of using a new CLYC crystal based detector in neutron porosity logging. MethodsThe Monte Carlo simulation method was employed to study the thermal neutron counting flux of 6Li CLYC detector with different purities and sizes. Thermal neutron counting of this new detector was compared with that of 3He detector under different pressures, the replacement of CLYC detector was verified according to the neutron porosity response of different detectors. ResultsThe results show that the new CLYC detector has high thermal neutron counting efficiency when its purity of 6Li and size are large. Compared with 3He detector, the CLYC detector with low thermal neutron detection efficiency has a near/far detector ratio similar to that of 3He detector after optimizing the combination of source diatances which can be effectively applied to obtain formation porosity. ConclusionsThis study provides a theoretical basis for the detector selection and replacement of neutron porosity logging tools in the future.
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Simulation study on COPRA corium pool based on COMSOL
Guangyu ZHU,Jinkun MIN,Li ZHANG,Yidan YUAN
Nuclear Techniques. 2021, 44 (4):
40601-040601.
DOI: 10.11889/j.0253-3219.2021.hjs.44.040601
BackgroundCOMSOL is considered to be an ideal approach to study the complicated physical processes in the corium pool because of its good simulation ability in multiphysics field. PurposeThis study aims to build a computational model by COMSOL to explore the turbulent flow field and phase change in the corium pool. MethodsBased on non-isothermal flow calculation module, the non-eutectic binary mixture of 20%NaNO3-80%KNO3 used in COrium Pool Research Apparatus (COPRA) experiment was simulated by using phase-change material model. The radiation heat transfer in the closed cavity at the top of corium pool was translated into the radiation heat transfer between the upper surface and the environment available as boundary condition in COMSOL. ResultsSimulation results show that in addition to the evident natural convection main flow, a mass of vortexes are existed in the corium pool, which lead to the thermal stratification of the corium pool in steady-state. Under the influence of natural convection main flow and gravity, the crust thickness along the cooling wall surface decreases with the vessel polar angle of corium pool increases. ConclusionsThe current computational model is helpful for severe accident mitigation system design.
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Study on the reinforcement process and performance of carbon fiber fabric for graphite control rod guide tube of molten salt reactor
Bo HUANG,Hong JIANG,Zhoutong HE,Fanggang LIU,Hui TANG,Xingtai ZHOU
Nuclear Techniques. 2021, 44 (4):
40602-040602.
DOI: 10.11889/j.0253-3219.2021.hjs.44.040602
BackgroundThe control rod guide tube (CRGT) is needed to facilitate the movement of the control rod in the molten salt reactor (MSR). Graphite material is resistant to neutron radiation, high temperature and molten salt corrosion, and has small neutron absorption cross section, hence it is an ideal CRGT material except for the comparatively low strength and toughness. It is believed that reinforcing the graphite CRGT with carbon fiber fabric (CFF) is a promising way to promote its application in molten salt reactor. PurposeThis study aims to optimize the preparation parameters of the CFF reinforcement for CRGT. MethodsFirst of all, the CFF was pre-impregnated with precursor solution in a vacuum environment, and wound on the graphite tube with a certain tension. Secondly, samples with different densification cycles and winding layers were prepared through multiple pressure impregnation, curing and carbonization cycles of the Precursor-Infiltration-Pyrolysis (PIP) process. Then, the mechanical properties of carbon fiber cloth reinforced graphite tube samples and the reference samples were tested. Finally, the microstructure and failure mode of the damaged samples were analyzed for further optimizing the preparation process to improve the material performance. Results & ConclusionsThe results show that the CFF winding and PIP method enhances the strength and toughness of the graphite tube, and the preparation parameters have been optimized that make the reinforced graphite CRGT with carbon fiber fabric (CFF) possible to be applied to MSR.
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Development and verification of fuel management code for liquid-fueled molten salt reactor based on deterministic code system
Kaicheng YU,Maosong CHENG,Zhimin DAI
Nuclear Techniques. 2021, 44 (4):
40603-040603.
DOI: 10.11889/j.0253-3219.2021.hjs.44.040603
BackgroundMolten salt reactor (MSR), a candidate of the Generation IV reactors, has the advantages of inherent safety and good neutron economy. The circulation flow of fuel, on-line refueling and reprocessing are the three features of liquid fuel in MSR. Due to the features of liquid fuel, the fuel management regulations for conventional reactors that use solid fuels are not applicable to liquid-fueled MSR. PurposeThis study aims to develope a new fuel management code named ThorNEMFM to analyze liquid fuel management of MSR. MethodsFirst of all, the self-developed deterministic code ThorCORE3D was coupled with the cross section processing module and a burnup calculation module. Then a fuel management analysis program ThorNEMFM was developed for liquid fuel molten salt reactor to realize on-line feeding and fission product processing. Finally, the ThorNEMFM was verified on molten salt reactor experiment (MSRE), molten salt breeder reactor (MSBR) and molten salt fast reactor (MSFR). ResultsThe results of benchmarks for MSRE, MSBR and MSFR indicate that ThorNEMFM has a high accuracy with benchmarks in comparison with the reference. ConclusionsThorNEMFM is suitable for calculation and analysis of fuel management of MSR.
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