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    20 November 2020, Volume 54 Issue 11
    Test of 235U Covariance Data with Nuclear Data Adjustment BenchmarkWU Haicheng
    WU Haicheng
    2020, 54(11):  1985-1991.  DOI: 10.7538/yzk.2020.youxian.0279
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    Nuclear data covariances are important input data for quantificational assessment of nuclear facility design uncertainty and nuclear data adjustment (NDA), which give direct impact on estimated uncertainties and results of NDA. To test the rationality of the newly evaluated nuclear reaction cross sections covariance in smooth region for 235U which is generated based on the analysis of source of experiments uncertainties and linear least-square method, the data were tested with the NDA benchmark exercises recommended by the OECD/NEA WPEC/SG33. The input data of the NDA generated from JENDL-4.0 library were updated with the cross sections and covariances of 235U fission, capture and inelastic scattering reactions from 235U cov, and used in NDA calculation. The new results were compared with the original JENDL-4.0 ones. The test results show that the covariance data from 235U cov are unreasonable. Too large uncertainties around threshold energy of inelastic scattering reaction were found. The uncertainties of fission and capture cross sections in smooth region are too small to be supported by NDA results.

    Boron Equivalent Measurement of Nuclear Graphite with Photoneutron Source
    WANG Xiaohe;HU Jifeng;CHEN Jingen;CAI Xiangzhou;WANG Naxiu;WANG Hongwei;HAN Jianlong
    2020, 54(11):  1991-1998.  DOI: 10.7538/yzk.2019.youxian.0781
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    Impurities in nuclear materials with high thermal neutron absorption cross section will change the reactivity. The absorption of thermal neutrons by these impurities is represented by boron equivalent, which is one of the important factors to measure the purity of nuclear materials. Boron equivalent can be determined directly via the measurement of macroscopic thermal neutron absorption cross section based on an isotopic neutron source, but with lower accuracy. The photoneutron source, which can generate neutrons with higher intensity, better direction and lower energy, can effectively improve the accuracy of boron equivalence measurement. Therefore, the boron equivalent measurement of nuclear graphite was carried out with the photoneutron source driven by 15 MeV electron LINAC. Monte Carlo simulation method was used to optimize the experimental scheme, and the experimental data were tested and modified. Finally, the quantitative analysis method was established for the measurement of graphite boron equivalent. This method can quickly and accurately measure the boron equivalent of nuclear materials, which is of great significance for the physical design and safety assessment of the reactor.

    Third Phase Formation in Extraction of Nd(Ⅲ) by (DdO)2DGA
    HE Xihong;LU Chun;YE Guoan;LI Linbo;CAO Zhi;WANG Xinyao
    2020, 54(11):  1999-2005.  DOI: 10.7538/yzk.2019.youxian.0789
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    N,N′-di-octyl-N,N′-di-dodecyl-3-oxapentane-1,5-diamide ((DdO)2DGA) is a new unsymmetrical diglycolamide extractant for the separation of trivalent actinides and lanthanides from high level liquid waste. In this paper, the representative element Nd in lanthanides was taken as the research object to study the third phase formation behavior of Nd(Ⅲ) in (DdO)2DGA, the influence of aqueous phase acidity, temperature and co-existing metal ion concentration on the limiting organic concentration (LOC) of Nd(Ⅲ) in (DdO)2DGA/n-dodecane solution was investigated. The structural variation of (DdO)2DGA-Nd(Ⅲ) complex formed in the organic phase before and after third phase formation was also studied by infrared spectra, UV-vis spectra and fluorescence spectra. The results show that LOC of Nd(Ⅲ) increases with temperature, and decreases with increase of aqueous acidity and co-existing metal ion concentration. However, the structure of (DdO)2DGA-Nd(Ⅲ) complex is not varied before and after third phase formation, and the reason for third phase formation is the variation of aggregation behavior of extracted complexes.

    Dissolution Behavior of Cerium in NaCl-KCl and NaCl-KCl-CeCl3 Molten Salt System
    ZHANG Lei;ZHENG Weifang;LIN Rushan;CHEN Hui;ZHANG Kai;SONG Wenchen
    2020, 54(11):  2006-2013.  DOI: 10.7538/yzk.2019.youxian.0904
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    The dissolution behavior of cerium in NaCl-KCl and NaCl-KCl-CeCl3 molten salt systems was evaluated by ion concentration determination, gas generation method, determination of acid consumption of molten salt, X-ray diffraction and differential gravimetry analysis method. The results show that the loss of cerium in molten salt involves physical dissolution and chemical reactions, which can produce oxides, nitrogen oxides, chlorine oxides and other forms. The cerium can also react with the composition of crucible. Cerium and its compounds have a low solubility in molten salts, but new phases are formed, which contain large amounts of cerium. The dissolution loss of cerium can be effectively reduced by using nested crucible and increasing the height of small crucible.

    Research Status of Geological Traceability Analysis of Uranium Ore Concentrate
    JIANG Xiaoyan;LI Lili;DU Zhiming
    2020, 54(11):  2014-2023.  DOI: 10.7538/yzk.2020.youxian.0364
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    In recent years, with the development of nuclear forensic analysis technology supported by high-accuracy and high-precision analysis methods, uranium isotopes, the distribution pattern of rare earth elements (REE), Sr and Nd isotope abundances, and some impurity element contents are mainly identified as effective geographic traceability fingerprint information of uranium ore concentrate (UOC). The main technological developments include different dissolution and chemical separation methods depending on the types of UOC samples, the advanced mass spectrometry technology used in the measurement of geographic characteristics such as the content and isotopic abundance of uranium and impurities in UOC, and the UOC standard samples with the same matrix as quality control (QC) measure used in the entire process of dissolution, separation and measurement. To promote the application of nuclear chemistry and geochemistry etc. in the study of nuclear forensics, based on the research achievements in recent years, the details of development in UOC’s nuclear forensic geographic traceability analysis technology and data processing are discussed.

    Flow and Heat Transfer Characteristics of Swordfish Fin Microchannel
    GONG Ya;GUO Zhangpeng;ZHANG Tianyi;WANG Shengfei;HUANG Yanping;NIU Fenglei
    2020, 54(11):  2024-2030.  DOI: 10.7538/yzk.2019.youxian.0642
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    The supercritical carbon dioxide Brayton cycle is a new generation of thermal cycle used in the fourth generation of nuclear energy. As a high or low temperature recuperator of supercritical carbon dioxide Brayton cycle, the thermal hydraulic characteristics of the compact microchannel heat exchanger directly affect the power cycle efficiency. Reducing flow resistance of the temperature recuperator while maintaining high heat transfer efficiency is an important research for microchannel heat exchanger optimization design. The swordfish fin microchannel design considering bionics theory can significantly reduce the flow resistance. In this work, the swordfish fin heat exchanger model was established with supercritical carbon dioxide fluid as the flow medium. The effect of swordfish fin design with different arrangements on heat transfer characteristics was analyzed by three-dimensional numerical simulation. At the same time, the thermal hydraulic characteristics of swordfish fin design were compared with those of traditional commercial Z-shaped microchannel heat exchanger. The results show that under the same Reynolds number, the Nusselt number of the swordfish fin microchannel is twice as much as that of the Z-shaped microchannel, but the pressure drop is only half of that. Therefore, the thermal hydraulic performance of swordfish fin microchannel heat exchanger is obviously better than that of Z-shaped heat exchanger. It is obtained from optimization analysis that the optimal pitches for swordfish fin design is that the La=8 mm along the flow direction, and the Lb=6 mm perpendicular to the flow direction.

    Experimental Study on Heat Transfer and Resistance Characteristics for Finned Tube of Fast Reactor Sodium-air Heat Exchanger
    ZHU Lina;YE Yuanwu;HOU Bin;CHEN Zhenjia
    2020, 54(11):  2031-2036.  DOI: 10.7538/yzk.2020.youxian.0173
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    Sodium-air heat exchanger is one of the important equipment in the decay heat removal system of sodium-cooled fast reactor, which together with the external environment constitutes the final heat trap for the residual heat of a reactor. Due to the vertically arranged finned tube structure adopted by the heat exchanger tubes of the sodium-air heat exchanger and the difference of the flow direction and velocity of the air at the different positions, the heat transfer and resistance characteristics are very different from of traditional finned tube heat exchanger. In this paper, based on the engineering design requirements of sodium-air heat exchanger, two kinds of experimental specimens were designed to study the heat transfer and flow resistance characteristics of finned tube bundles with air flow angles of 90° and 30°. The experimental results show that the heat transfer and resistance coefficients of the same finned tube with air flow angles of 90° are significantly greater than those of the finned tube with air flow angles of 30°. For the same air flow direction, the second finned tube has the largest heat transfer coefficient. The research provides a theoretical basis for the design and optimization of sodium-air heat exchanger.

    Numerical Simulation on Heat Transfer Characteristic of Steam Generator of ACME Facility
    LI Weiqing;LYU Yufeng;ZHAO Minfu;ZHONG Jia;WANG Nan;ZHANG Peng
    2020, 54(11):  2037-2044.  DOI: 10.7538/yzk.2019.youxian.0832
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    Taking the steam generator (SG) of the ACME facility as the study object, two-fluid model was chosen for the secondary side of the SG and a direct simulation was made on the whole SG of the ACME facility with the CFD software CFX. The stable test conditions were calculated. The primary side and secondary side temperature distribution of SG, the secondary side void fraction distribution and the wall temperature of heat transfer tube were obtained. The secondary side detailed flow and heat transfer characteristics were arrived. The results show that bubbles begin to accumulate from the second baffle and bubble amount increases as the tube is higher. Near the bend area and above, they all become steam. The calculated results are all in accordance with the experimental results.

    Development of System Code FR-Sdaso for Sodiumcooled Fast Reactor
    WANG Xiaokun;QI Shaopu;YANG Jun;YE Shangshang;WANG Lixia;FENG Zongrui;CHONG Daotong;JIA Hongyu;YANG Xiaoyan;LIU Yizhe;YANG Hongyi
    2020, 54(11):  2045-2053.  DOI: 10.7538/yzk.2020.youxian.0306
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    To solve actual problems in the accident analysis and working condition design of the 600 MW demenstration fast reactor (CFR600), the sodium-cooled fast reactor (SFR) system code FR-Sdaso was developed, which could be used to model the reactor core, primary system, secondary system, tertiary system, quadruple system and the decay heat removal system of the SFR. The physical models can be divided into three categories: The models for nuclear island equipment including point reactor model, single-channel core thermal model, multi-zone sodium pool model and four-zone steam generator model, etc., the lump parameter models for conventional island equipment, including turbine, condenser, feed water heater, deaerator, etc., and the general models for pump, valve, pipe and control volume. Preliminary V&V work for FR-Sdaso was conducted, and the results show that FR-Sdaso can be used to analyze the transient conditions of the whole plant and typical SFR accidents such as overpower, loss of flow, and loss of heat sink. FR-Sdaso was used in the design and safety analysis of the CFR600.

    Uncertainty Analysis on Boundary Condition in Subcooled Boiling Flow by Deterministic Sampling
    ZHANG Xiang;PENG Minjun;CONG Tenglong;LI Xiaojia;CHEN Yiran
    2020, 54(11):  2054-2062.  DOI: 10.7538/yzk.2019.youxian.0782
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    In this work, the DEBORA subcooled boiling flow experiments were employed as benchmark and simulated by Fluent. The uncertainty from boundary conditions and its influence trend of the prediction of subcooled boiling flow were analyzed by using deterministic sampling method. The distributions of radial parameters were calculated and the confidence intervals were obtained. Besides, the uncertainty source contributions on radial parameters were also analyzed. The results show that the uncertainties of the inlet temperature and the wall heat flux play predominant roles on the prediction of radial void fraction, while the uncertainties of the operating pressure and the inlet temperature dominate calculation of radial liquid temperature.

    Research on Coupling of Neutronics and Thermal-hydraulics for Fuel Assembly of Thorium Molten Salt Reactor Moderated by Zirconium Hydride Rod
    ZHU Fan;WU Jianhui;YU Chenggang;MA Yuwen;CHEN Jingen;CAI Xiangzhou
    2020, 54(11):  2063-2072.  DOI: 10.7538/yzk.2019.youxian.0803
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    Based on Monte Carlo particle transport code MCNP and self-developed sub-channel thermal-hydraulic code SubTH, a code system MCNP-SubTH coupling neutronics with thermal-hydraulics was developed, which was suitable for steady state analysis for a thorium molten salt reactor moderated by zirconium hydride rod (ZrH-MSR). It solved the difficulties in the neutronics and thermal-hydraulics coupling code due to different mesh types, and has a general validity. MCNP-SubTH exchanged data between MCNP and SubTH by an external coupling. The power density field obtained from MCNP was provided as a SubTH solution file to give a user-specified source term, and then the density and temperature field from SubTH was updated and as a new MCNP input file by MCNP-SubTH to realize iterative calculation. The accuracy of MCNP-SubTH was verified by each relatively independent module. MCNP-SubTH application in the fuel assembly of ZrH-MSR was studied, and its validity was verified.

    Analysis of Effect of SBLOCA before and behind Isolation Valve of PRHRS on Passive Reactor Core Cooling System
    HAO Botao;WANG Nan;ZHONG Jia;SHI Yang;FANG Fangfang
    2020, 54(11):  2073-2080.  DOI: 10.7538/yzk.2019.youxian.0877
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    For the test process of small break loss of coolant accident (SBLOCA) of AP type nuclear power plant, there is a more consistent understanding at home and abroad. However, the influence of the same size of the break on the test process and passive core cooling system in different locations still needs further study. In this paper, a large passive core cooling integrated test facility ACME was used to study the break accident test conditions of passive residual heat removal system (PRHRS) before and behind the isolation valve, and the bottom break test of the cold pipe of core makeup tank (CMT) was used as the contrast condition. The test results show that the accident process of PRHRS before and behind the isolation valve is in accordance with the process of SBLOCA, the core is always in a good cooling statement and the safety of passive core cooling system is effectively verified. There is a certain impact on the accident process for the same break size and different break locations, and the location of the break has a key impact on the CMT level and safety injection flow. In contrast, the heat transfer of PRHRS equipment is also quite different. The heat transfer of cold pipe break and break behind the isolation valve is greater than break before the isolation valve, however, the flow and heat transfer mechanism of PRHRS heat exchange tube needs further study.

    Transient Safety Analysis of Small Leaded-Bismuth Cooled Fast Reactor
    ZHANG Yifan;LIU Zhouyu;CAO Liangzhi;ZHENG Youqi;SHAO Yiqiong
    2020, 54(11):  2081-2088.  DOI: 10.7538/yzk.2019.youxian.0812
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    Based on the multi-channel thermal model and the power model, the calculation code which could be used in the transient safety analysis of fast reactor was developed in unprotected overpower accident, unprotected loss of flow accident and unprotected loss of hot sink accident in the paper. By this code, the core reactivity, power and thermal parameter changes with time in different accident cases were calculated and the core neutronics and thermal-hydraulics performance was analyzed. The results indicate that the core design has safety features when accident happens.

    Transient Safety Characteristic Analysis of Xi’an Pulsed Reactor during Reactivity Insertion Accident
    TIAN Xiaoyan;CHEN Sen;YANG Ning;ZHU Lei;LI Huaqi;MA Tengyue;HU Pan;KANG Xiaoya
    2020, 54(11):  2089-2097.  DOI: 10.7538/yzk.2019.youxian.0854
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    In order to study the transient safety characteristics of Xi’an Pulsed Reactor (XAPR) when unexpected reactivity insertion accident happened and shutdown system failed, the main mathematical models were established based on the specific core structure and operation conditions of XAPR. Meanwhile, a transient thermal-hydraulic code called TSAC-XAPR was developed to analyze the safety characteristics of XAPR. The TSAC-XAPR code was then used to simulate the reactivity insertion accident of XAPR. The calculation results indicate that when XAPR operating under rated power, reactor can reach a new steady state for reactivity insertion accident, depending on its inherent feedback mechanism. When XAPR operating under high power, especially above the critical power, key thermal-hydraulic parameters of reactor will tend to oscillate and can’t reach a steady state again for reactivity insertion accident. Besides, it is also found that different reactivity insertion modes will only affect the variation trend during the phase of reactivity insertion instead of the final value at steady state.

    PSA of CPR1000 SBO Accident Based on Monte Carlo Accident Process Analysis
    WANG Zhao;YANG Jianfeng;FENG Bingchen
    2020, 54(11):  2098-2106.  DOI: 10.7538/yzk.2019.youxian.0707
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    A mathematical model for the process of possible accident sequence in the CPR1000 nuclear power plant was established, and the Monte Carlo method was used to programming codes to calculate the possibility of timely recovery of AC power in each accident process. According to the calculation results, probabilistic safety analysis (PSA) of the station black out (SBO) was conducted in the paper. The results show that using Monte Carlo method to analyze the process of SBO can make the PSA more in line with the actual situation of the nuclear power plant. And the overall risk of the nuclear power plant and the risk of SBO can be understood better by using the method.

    Design of Test Scheme for Failure Rate Verification of Control Rod Drive Mechanism Drop
    XIAO Lili;GU Jipin;JIN Fenglei;PU Enshan;LIU Xiuting
    2020, 54(11):  2107-2112.  DOI: 10.7538/yzk.2020.youxian.0032
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    In the event of the reactor accident, the control rod drive mechanism (CRDM) must be able to drop quickly and achieve the safe shutdown. In order to ensure the safety of the reactor, the nuclear power plant puts toward higher requirements on the failure rate of the CRDM. In this paper, the design algorithm of the test scheme for the failure rate of the CRDM and the corresponding solution method based on the dichotomy method were proposed. The correctness was verified by comparison with the scheme given in GB 5080.5-85. It supplements the design algorithm of the fixed-end censored test scheme  for success rate verification test scheme of the success or failure type products in GB 5080.5-85. Based on any given success rate, risk value and discrimination ratio, it can give the censored test protocol in a short time. It also designs the success rate verification test scheme for general success or failure type products to meet the increasing reliability requirements for success or failure type products.

    Sodium Fire Accident Scenario Simulation Technology Study Based on FDS Code
    YANG Hongyi;SONG Wei
    2020, 54(11):  2113-2120.  DOI: 10.7538/yzk.2019.youxian.0918
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    Sodium fire is a typical accident of sodium-cooled fast reactor. Simulation of sodium fire accident scenario by software is a powerful tool for risk assessment of sodium fire accident. In this paper, the conventional fire three-dimensional computational fluid dynamics software FDS was used as a platform to add a sodium fire combustion model, including combustion heat model, combustion rate model, spray liquid sodium particle size distribution model, etc., and complete the development of sodium fire scenario modeling analysis program. And through the comparison with SPHINCS sodium fire test and calculation results, the feasibility of the method and development plan was verified. The research results of this paper can provide the research basis and experience reference for the development of the subsequent sodium fire simulation program.

    Application and Research of Super Homogenization Method for PWR Core Pin-by-pin Calculation
    ZHANG Bin;LI Yunzhao;WU Hongchun;WANG Dongyong;LIU Yong;YU Yingrui
    2020, 54(11):  2120-2126.  DOI: 10.7538/yzk.2019.youxian.0924
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    Super homogenization (SPH) method was used as the homogenization technique of PWR core Pin-by-pin calculation. For the fuel assembly, the traditional SPH method was used to generate the group constants. For the reflector-assembly, the spatial leakage dependent SPH method was evaluated. The problem of non-convergence in the iteration of SPH factors calculation was solved by improved SPH method which can keep the neutron leakage and reaction activity conserved. Based on KAIST benchmark, the performance of the SPH method applied in PWR core Pin-by-pin calculation was evaluated. Numerical results demonstrate that compared with the results of traditional assembly-homogenized calculation, the Pin-by-pin calculations have the higher accuracy.

    Exponential Short Characteristic Spatial Discretization Method for SN Transport Calculation on 3D Structured Grids
    LIU Cong;HU Xiaoli;ZHANG Bin;CHEN Yixue
    2020, 54(11):  2127-2136.  DOI: 10.7538/yzk.2019.youxian.0891
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    Proper spatial discretization methods and mesh sizes are crucial for the computational accuracy of the discrete ordinates methods applied on the particle transport simulations. Short characteristic spatial discretization splits the computational cell based on the relationship between incoming and outgoing particles, and is capable of mitigating non-physical oscillations caused by the deficiency of the spatial variable discretizing. Exponential short characteristic (EC) is suitable for shielding problems with deep-penetration and strong attenuation, but there are expensive numerical processes including the redundant geometric calculation and nonlinear solving in the previous studies. A modified EC scheme was developed based on 3D structured grids, and the projection mapping was avoided by applying the new splitting-substituting scheme. Optimized exponential moment and exponential factor solving algorithms improve the computational efficiency. From results of four fixed-source problems with simple geometry, accuracy and efficiency were compared among several spatial schemes. The accuracy of EC on the coarse meshes is much better than that of other methods, and its excellent coarse mesh accuracy is cable of compensating the expensive computational cost. Proper mesh distributions help to realize the EC’s advantage of the computational efficiency.

    Development and Validation of Burnup Code Based on High-order Chebyshev Rational Approximation Method
    ZHANG Binhang;YANG Senquan;CHEN Yunlong;HONG Feng;YUAN Xianbao;ZHANG Yonghong;TANG Haibo;FENG Hongying
    2020, 54(11):  2137-2144.  DOI: 10.7538/yzk.2019.youxian.0913
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    Based on high-order Chebyshev rational approximation method (CRAM), a point-burnup code named ICRAM was developed and internally coupled to Monte Carlo code OpenMC, forming a burnup calculation and analysis program OPICE. Compared with the traditional partial fraction decomposition (PFD) form of CRAM, the high-order incomplete partial fractions (IPF) form of CRAM has the characteristics of good numerical stability, high calculation accuracy and better step tolerance, etc., which meets the needs of high-fidelity burnup calculation development. In order to improve the accuracy of coupling calculations, two coupling strategies including prediction-correction method and sub-step method were implemented in OPICE. Three different calculation modes were supported by OPICE to execute the decay, constant flux and constant power calculations. By calculating the OECD/NEA burnup benchmark and fast reactor burnup benchmark, the calculation results of OPICE are in good agreement with the experimental data and each reference value. The correctness and validity of OPICE are verified preliminarily.

    Boundary Condition Processing Method for MOC Calculation in Arbitrary Geometry
    YANG Rui;HU Yun;SHAN Haodong;XU Li
    2020, 54(11):  2145-2152.  DOI: 10.7538/yzk.2019.youxian.0849
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    Boundary condition processing is one of the difficulties encountered in the application of method of characteristics (MOC) to arbitrary three-dimensional geometry. In this paper, a boundary condition processing method was proposed, which not only preserved the track continuity as cyclic track method, but also could be applied to arbitrary geometry as the interpolation method. The MOC equation was derived under the flat source approximation and an internal iterative method was proposed in which the source term and the boundary angular flux were processed separately. It was proved that the equation had a unique solution which could be constructed similarly to the cyclic track method. The iterative calculation flow was given by numerical integration and weight interpolation. Takeda benchmark, single uranium sphere model with water cavity and C5G7 benchmark were calculated to test the accuracy. The maximum error of keff is 21, 319 and 138.8 pcm respectively, which shows that the method is reliable. This method can be applied to arbitrary geometry without storing boundary fluxes and performing boundary iteration.

    Application of High Performance Computing Analysis Technology in Shielding Design of Large Sodium-cooled Fast Reactor
    WANG Shixi;WU Mingyu;ZHANG Qiang;WANG Fenglong;WANG Yi;SHAO Jing;WAN Haixia;FU Yuanguang;CHENG Tangpei
    2020, 54(11):  2153-2158.  DOI: 10.7538/yzk.2019.youxian.0892
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    The sodium-cooled fast reactor container is an integrated pool structure composed of numerous internal components and complex structure. The anisotropy is obvious and the deep penetration problem is serious in the process of neutron transport from core to biological shielding. The calculation of three-dimensional SN method in large scale is the bottleneck restricting in the design of fast reactor shielding. By combining with high performance computing technology, the parallel computing scheme is used to solve the anisotropic three-dimensional deep penetration shielding calculation  in the fast reactor. In this paper, the China Demonstration Fast Reactor (CFR600) reactor block was taken as the research object. Using JSNT-CFR code, the neutron flux rate, photon flux rate, and dose rate in the reactor block were calculated in detail. The calculation results were compared with those of the existing two-dimensional code. The results show that combining the traditional shielding calculation method with high performance computing can meet the requirements of CFR600 reactor block shielding calculation accuracy, and obtain a more comprehensive three-dimensional display effect. It can solve the problem of shielding calculation of complex problems such as complex model and particle penetration depth. It has obvious advantages and provides strong support for the large sodium-cooled fast reactor shielding design.

    Effect of Heat Treatment in α+β Phase Field on Microstructure and Corrosion Property of Zr-Sn-Nb Alloy
    JIA Yuzhen;DAI Xun;WANG Pengfei;LIU Hong;PENG Qian
    2020, 54(11):  2159-2165.  DOI: 10.7538/yzk.2019.youxian.0823
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    Uniform corrosion tests were carried out with the specimens prepared by different heat treatments at the temperature in α+β phase field. The surface microstructure of specimens was observed by  scanning electron microscope, the corrosion behavior was analyzed by autoclaves, and the oxide layer on the surface after the corrosion test was analyzed by focused ion beam (FIB) and atomic force microscope (AFM). The results show that after the heat treatment in α+β phase field, lamellar β-Zr phase appeares in the Zr matrix, and after the subsequent α phase final heat treatment, the β-Zr phase will be decomposed to α-Zr and discontinuous second phase particles. For the specimens heat treated in α+β phase field, after the α phase final heat treatment, the corrosion resistance under 360 ℃/18.6 MPa pure water condition is better than that of specimens without final heat treatment. The oxide film formed on the β-Zr protrudes on the oxide surface, on the contrary, after α phase final heat treatment, β-Zr decomposes, and the oxide layer is sunken in this area.

    Microstructure and Corrosion Behavior about Zr-Cu-Cr Alloy
    ZHANG Ge;LI Qiang;PENG Jianchao;LIANG Xue;PENG Liting;YAO Meiyi
    2020, 54(11):  2166-2173.  DOI: 10.7538/yzk.2019.youxian.0863
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    Seven kinds of Zr-Cu-Cr alloy samples were prepared by using crystalline zirconium as the base material, through normalized processing and annealing at 600 ℃/5 h. The corrosion tests were carried out in static autoclave with different water chemistry conditions, and the microstructures of alloys were studied by using EBSD, SEM and TEM to investigate the effects of Cu and Cr interaction on the microstructure and corrosion resistance of zirconium alloys. The results show that the addition of Cu element refines the recrystallized grains. When the Cr content is 1.0%, it has grains with a size of 40 μm or more. There are two kinds of second phases in Zr-Cu-Cr alloys. Zr2Cu phase (bct) with a size of 100 nm or more and ZrCr2 phase (hcp) with a size of 60 nm or less. As the content of Cu increases, the number of Zr2Cu increases. As the content of Cr increases, the size of ZrCr2 phase does not change obviously, but the number and distribution band density increase. When exposed to the superheated steam at 400 ℃/10.3 MPa, the corrosion resistance of Zr-0.3Cu-0.2Cr and Zr-0.3Cu-0.5Cr alloys is worse, the other alloys still don’t have corrosion transition after corroding 100 days, and the corrosion resistance is better. Zr-1.0Cr alloy has the best corrosion resistance. When exposed to the 360 ℃/18.6 MPa/0.01 mol/L LiOH aqueous solution for 42 days, all of alloys have poor corrosion resistance, and the addition of Cu element reduces their corrosion resistance.

    Effects of Alloy Elements on Irradiation Embrittlement of Reactor Pressure Vessel Steel: Insights from Neural Network Modelling
    JIA Lixia;HAN Xu;BAI Bing;WANG Dongjie;YANG Wen
    2020, 54(11):  2174-2181.  DOI: 10.7538/yzk.2019.youxian.0856
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    As one of the key components that can not be replaced in PWR, the safety and stability of reactor pressure vessel (RPV) steel determine the safety and economy of the reactor. The irradiation embrittlement of RPV steel is the limiting factors for the operation of PWR. The irradiation embrittlement of RPV steel is closely related to its alloy composition. Based on the machine learning method, the relationship between key alloy components (Cu/Mn/Ni/Si/P) and irradiation embrittlement of RPV steel was constructed. The results show that the relationship between the alloy composition and irradiation embrittlement is basically consistent with the traditional cognition. The irradiation embrittlement is sensitive to Cu content, and Cu-Ni has synergistic effect on irradiation embrittlement. In low Cu alloys, Mn-Ni and Ni-Si have synergistic effects on embrittlement.

    Effect of Coating and Heat Treatment on High Temperature Burst Property of Stainless Steel Thin-walled Tube
    HAN Zhibo;YANG Hongguang;ZHANG Jiantong;YUAN Xiaoming;LIU Shanshan
    2020, 54(11):  2182-2187.  DOI: 10.7538/yzk.2020.youxian.0316
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    In order to investigate the high temperature burst behavior and effect of coating and heat treatment on burst properties of stainless steel thin-walled tube for pressurized water reactor (PWR) and supercritical water-cooled reactor (SCWR) under the simulated LOCA (loss of coolant accident) conditions, the transient-heating burst tests of original, heat treatment and coated tube specimens were carried out under the simulated LOCA conditions by using the high temperature burst facility. The data of high temperature burst properties based on three kinds of 316L stainless steel tubes and the relationships between burst strength and total circumference elongation (TCE) with temperature were obtained under the conditions of 600-1200 ℃ and heating rate of 5 ℃/s, meanwhile, the fracture morphology and microstructure of burst test specimens were analyzed. The results of the burst test show that heat treatment temperature of coating preparation is the main reason for the decrease in burst strength, and the coating has little effect on the high temperature burst strength.

    Simulation Research on Control System of MNSR
    HONG Jingyan;PENG Dan;HAO Qian;WU Xiaobo;LI Yiguo;WANG Mengjiao
    2020, 54(11):  2188-2193.  DOI: 10.7538/yzk.2020.youxian.0309
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    By the theoretical analysis of the Miniature Neutron Source Reactor (MNSR) control system, the corresponding Simulink model was established and the simulation analysis of the MNSR control system was carried out. To facilitate the stability analysis, the mathematical model of the MNSR control system was reduced and discretized, based on the open-loop transfer function after reduction and discretization, the stability analysis was performed using the rltool analysis toolbox of the Matlab, and the critical gain for different sampling time was obtained. The calibration of the proportional coefficient Kp for the PID controller at different sampling time Ts was carried out with the introduction of positive step reactivity in the steady-state operation of the reactor, and the optimal calibration values of Kp at different sampling time were determined. The optimal differential coefficient Kd at different Ts and Kp was calibrated. After calibrations of TsKp and Kd, the reactor neutron flux density, control rod velocity and response to control rod introduction reactivity at negative step reactivity input and slope reactivity input were analyzed at Ts=60 ms, Kp=2 500 and Kd=300. The results of the simulation analysis show that the Simulink modeling about the MNSR control system is accurate and the analytical data are reliable, which provides a theoretical basis for software and hardware designs of the MNSR control system.

    Root Cause Analysis of Vibration Exceeding and Design Optimization of Stagnant Branch of Reactor Coolant Loop
    CAI Kun;ZOU Jianrong;CAI Yilin;MA Zhicai;ZHU Changfan;QIU Jian;ZHENG Mingguang
    2020, 54(11):  2194-2200.  DOI: 10.7538/yzk.2020.youxian.0194
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    During the hot function test of nuclear power plant, it was found that the vibration of a stagnant branch of reactor coolant loop (RCL) exceeded the limit. According to the natural modal analysis, acoustic modal analysis and the variation trend of measured vibration with temperature, it is inferred that the acoustic vibration of the fluid in the pipeline causes the resonance of the pipeline. According to the calculation of vortex shedding frequency and acoustic vibration frequency, it is inferred that the acoustic vibration and the vortex shedding frequency of fluid at the tee are locked, which leads to the amplification of acoustic vibration excitation. Based on the analysis result and measured data above, it is determined that the cause of the vibration exceeding is the fluid-acoustic-structure coupling vibration. The pipe supports were optimized to  avoid resonance frequency of acoustic structure coupling. The fillet at the tee was used to weaken the vortex shedding, so as to reduce the amplitude of flow-excited acoustic resonance. The vibration of the optimized branch will be significantly reduced to ensure the safety of the pressure boundary.

    Numerical Investigation on Temperature Characteristic of Capsule Surface and Filling Tube of Cryogenic Target under Radiation Condition
    GUO Fucheng;LI Cui;CHEN Guanhua;LI Yanzhong
    2020, 54(11):  2201-2208.  DOI: 10.7538/yzk.2019.youxian.0790
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    During the ice-laying period, a phenomenon is observed that the ice crystal could not be maintained in the filling tube, which results in the direct connection between the capsule and the deuterium source at high temperature. In this paper, a 3D cryogenic target model was established to study the influence of several factors on the temperature along capsule surface and filling tube. The results show that changing the transmittance of the sealing film can effectively solve the problem of being unable to block the filling tube, while changing the shield temperature and the surface emissivity of the aluminum enclosure has no obvious effect on that problem. It is found that the crystal can be maintained in the filling tube under the boundary conditions discussed in this paper with the transmittance of the sealing film greater than 0.025.

    Crystal Growth Technology of Deuterium-deuterium in Cryogenic Target Based on Temperature Shock
    TAO Chaoyou;YANG Hong;DAI Fei;LIN Wei;WANG Kai
    2020, 54(11):  2209-2216.  DOI: 10.7538/yzk.2019.youxian.0804
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    The quality of deuterium-deuterium (D2) ice layer in inertial confinement fusion cryogenic targets plays a crucial role in the success of fusion experiments. At present, the methods reported in the literature for preparing D2 ice layer in cryogenic target do not have good operability, and the technology and process are indeterminate, which restricts the formation of high-quality ice layer. Therefore, in this paper, the technology of the combination of temperature gradient, cooling rate and temperature shock ice was used, and on the basis of the control technology of residual ice, the controllable growth of high-quality ice layer was realized. Meanwhile, the effect of temperature control on the crystallization quality of D2 fuel in target pellets and the crystal growth of process of D2 fuel at cryogenic temperature were studied. The crystallization behavior of D2 fuel was analyzed using the crystal growth morphodynanics theory. Analysis of the bright ring in the backlit shadow image shows that D2 ice layer uniformity is 85.2%, the thickness is 40.35 μm, and the inner surface roughness is 2.15 μm. This method broadens the technology of the control of D2 ice seed and crystal growth at cryogenic temperature, lays a solid foundation for DT layer, and forms a certain technical reserve.

    Monte Carlo Simulation of Dose Calculation Parameter of Radioactive 125I Brachytherapy Source
    LIU Chuanfeng;SONG Mingzhe;WANG Hongyu;NI Ning;WEI Kexin;LIU Yuntao
    2020, 54(11):  2217-2222.  DOI: 10.7538/yzk.2019.youxian.0868
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    The end of silver rod of the domestic 125I brachytherapy source is right angle type, which is slightly different from the typical model of 6711 125I brachytherapy source. And it can have influence on the dose calculation parameters. Based on the structure of domestic 125I brachytherapy source, dose calculation parameters which are recommended by AAPM TG43-U1 were calculated by Monte Carlo method. The influence of the end of silver rod on the dose calculation parameters was studied. The simulation result of dose rate constant is 0.955 cGy·h-1·U-1 when the air kerma strength was calculated by the point detector, and the difference with the result of the TG43-U1 is within 1.03%. The radial dose function g(r) in the range of 0.05-10 cm at the transverse axis was calculated precisely. Then empiric equation was acquired by curve fitting. 2D anisotropy function F(r,θ) was calculated in 0°-90° and 0.25-7 cm. The source of the right angle structure of the end of the silver rod would cause a hump area of 2D anisotropy function when r equals 0.25 cm.

    Linear Coefficient of Thermal Expansion and Mechanical Property of Pb/B4C/BPR Composite
    MA Jiaxu;LIU Ying;LI Jun;LI Cheng;MA Lingcheng
    2020, 54(11):  2223-2231.  DOI: 10.7538/yzk.2019.youxian.0880
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    Lead/boron carbide/boron phenolic resin (Pb/B4C/BPR) is a new kind of shielding composite with high temperature resistance and high shielding performance for radiations. In this paper, the effects of lead powder content, particle size and surface treatment (SiO2 coating) on the linear coefficient of thermal expansion (CTE) and mechanical properties of the composites were systematically studied. The results show that the composites have the smallest linear CTE and excellent comprehensive mechanical properties when the lead powder content is 36% (volume fraction) and the particle size is 38 μm. Proper coating of SiO2 on the surface of lead powder effectively reduces the density difference between lead powder and B4C powder, which not only significantly reduces the linear CTE of the composites, but also improves the tensile, flexural strength and impact toughness. However, too much SiO2 coating content will result in the linear CTE of the composites increasing and the tensile, flexural strength and impact toughness reducing.

    Design and Recent Progress of High Intensity Accelerator for Jinping Underground Nuclear Astrophysics Experiment (JUNA)
    CHEN Lihua;CUI Baoqun;MA Yingjun;MA Ruigang;TANG Bing;HUANG Qinghua;MA Xie;LIAN Gang;GUO Bing;LIU Weiping;SUN Liangting;WU Qi
    2020, 54(11):  2231-2237.  DOI: 10.7538/yzk.2019.youxian.0865
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    Direct measurement of the cross sections for the key nuclear reactions is crucial for obtaining benchmark data for stellar model, verifying extrapolation model, constraining theoretical calculations, and solving key scientific questions in nuclear astrophysics. However, cross sections of the astrophysical reactions are extremely small. Tiny reaction rates in laboratories at the earth surface are hampered by the cosmic-ray background. With the ultra-low background, underground lab becomes a promising solution of experimental nuclear astrophysics. China Jinping Underground Laboratory (CJPL) is currently deepest underground site in the world. For such experiments, a 400 kV and 10 mA accelerator specially designed for Jinping Underground Nuclear Astrophysics Experiment (JUNA) will be placed in CJPL. In this paper, JUNA high intensity accelerator was introduced. Its layout, design considerations, beam optics and recent progress in ground laboratory were presented.

    3D Position Information Extraction of Refilled Charge during Injection Process of Electron Storage Ring
    ZHOU Yimei;LENG Yongbin;ZHANG Ning;GAO Bo;CHEN Zhichu
    2020, 54(11):  2238-2244.  DOI: 10.7538/yzk.2019.youxian.0867
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    The injection process of electron storage ring is a special transient process. The mismatch between the injector and the storage ring can be evaluated online by studying the 3D merging of refilled charge and stored charge. For the next-generation light source, it can lay a technical foundation for building a diagnostic system. The beam instrumentation group of the Shanghai Synchrotron Radiation Facility developed a 3D bunch-by-bunch diagnostic system based on button pick-up signal and high-speed acquisition boards to measure the beam transverse position and longitudinal phase. Extracting the 3D position information of the refilled charge during the injection process is the critical issue in this system. The charge-weighted average method was proposed to extract the transverse position and the proportional coefficient method was proposed to extract the longitudinal phase. The transverse betatron oscillation and longitudinal synchrotron oscillation of the refilled charge were analyzed. 3D position information of the refilled charge and several dynamic parameters of ring, such as longitudinal maximum oscillation amplitude, initial arrival time and synchrotron damping time, can be extracted during the user operation mode, which provides a strong toolkit for accelerator physics.

    Development of Thin-wall Vacuum Chamber for Heavy Ion Medical Accelerator
    LUO Cheng;YANG Weishun;XIE Wenjun;LI Changchun;CHAI Zhen;ZHU Xiaorong;LIU Jianlong;JIAO Jiqiang;WAN Yapeng;LIN Xiaojian;MA Xiangli;ZHANG Xiping;MENG Jun;CHEN Shuping
    2020, 54(11):  2245-2251.  DOI: 10.7538/yzk.2019.youxian.0798
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    Heavy Ion Medical Machine (HIMM), developed by the Institute of Modern Physics, Chinese Academy of Sciences, is the first medical heavy ion accelerator with independent intellectual property rights in China. Because the RAMPING mode was used for high frequency pulse dipole magnets of HIMM and the rising rate of magnetic field is 1.6 T/s, the vacuum chamber installed in the high frequency pulsed magnet is a thin-wall stainless vacuum chamber with reinforcing ribs to reduce the influence of eddy current on the ion beam stability. However, the gap size of magnet occupied by thin-wall stainless vacuum chamber with reinforcing ribs is too large, and it not only causes the high cost of magnets, but also greatly improves the maintenance cost. Based on these reasons, a new thin-wall vacuum chamber (0.3 mm) with ceramic lining was put forward and the prototype was designed and manufactured. The test results show that the obtained pressure of the prototype is in the order magnitude of 10-10 Pa, and the magnet gap can be effectively reduced. And it is the development direction of thin-wall vacuum chamber of accelerator in the future.

    Design of Fast Corrector Magnet Power Supply for HEPS Storage Ring
    LIU Peng;LONG Fengli;LI Yang;HAN Chao
    2020, 54(11):  2252-2257.  DOI: 10.7538/yzk.2019.youxian.0843
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    The high energy photon source (HEPS) has an ultra-low emittance of 0.1 nm·rad, which puts forward high requirement for beam stability. A high performance fast corrector magnet power supply was developed to improve the performance of the fast orbit feedback system to improve the beam stability. The prototype adopted a novel multi-level cascade topology structure to improve the dynamic performance, and a novel control method based on digital control was used to improve the dynamic performance of the prototype. Software simulation and the prototype test data show that the prototype has a small signal bandwidth of 10 kHz, step response time lower than 70 μs, output current ripple of less than 20 ppm. The power supply prototype can meet the requirement of HEPS.

    Simulation and Performance Study on Compact Thermal Neutron Collimator Based on Gadoliniumdoped Silicate
    LI Da;LIU Yongde;YANG Lifang;GUO Liang;LIU Likun;ZHOU Zhibo
    2020, 54(11):  2258-2263.  DOI: 10.7538/yzk.2020.youxian.0337
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    In order to achieve the goal of nuclear security, the neutron generator is used to image shielded nuclear material by non-destructive analysis method. In order to improve the imaging quality, it is necessary to collimate the slowed neutrons. In this paper, the Monte Carlo software Geant4 was used to theoretically model a new compact thermal neutron collimator based on gadolinium-doped silicate, and the thermal neutron transmittance and collimation ratio of the collimator were simulated and calculated. The calculation results will be used to guide the subsequent construction of neutron-based nuclear material imaging systems.

    Location Method of Radioactive Source Based on Improved Particle Filtering
    LIU Haojie;XIAO Yufeng;ZHANG Hua;TIAN Xinghao;ZHANG Cheng
    2020, 54(11):  2264-2272.  DOI: 10.7538/yzk.2019.youxian.0753
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    In order to use the autonomous mobile robot to search for unknown radioactive objects, a method for locating the radioactive source with improved particle filtering was proposed based on the recursive Bayesian estimation model. Firstly, an initial particle set was established, and the weight of the particle was updated and normalized according to the observation value. Secondly, the auto-optimal resampling method was introduced to increase the particle diversity in the resampling process. Finally, the particle carrying out the convergence condition was weighted and summed to estimate the position and activity parameters of the radioactive source. The simulation results show that the method is feasible and effective. The positioning accuracy is high in the unshielded environment, and the approximate position of the radioactive source can also be found in the shielded environment, which provides a reference for the final location of the radioactive source.

    Modelling of Cellular Survival Following Chromosome Aberration
    WANG Wenjing;LI Chunyan;QIU Rui;WU Zhen;ZHANG Hui;LI Junli
    2020, 54(11):  2273-2281.  DOI: 10.7538/yzk.2019.youxian.0806
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    In order to assess the biological effectiveness of protons and heavy ions, a mechanistic model of cellular survival following radiation-induced chromosome aberrations was proposed. DNA damage repair module of nanodosimetry biophysical Monte Carlo simulation code (NASIC) was developed to simulate chromosome aberrations in nucleus induced by different types of particles at different linear densities of energy transfer (LET). Based on the analysis of radiation-induced chromosome aberration, a mechanistic model of cellular survival following radiation-induced chromosome aberrations was proposed. The parameters of the model could be obtained by fitting the experimental data of cell survival curves. For V79 cells exposed to X-rays and protons at different LET, the cell survivals calculated with the proposed model are in good agreement with the experimental data, and the correlation coefficient is 0.985 3. The proposed model was validated by experimental data of V79 cells exposed to 4He ions and the correlation coefficient of the cell survival calculated with the proposed model, and the experimental data is 0.931 1. The results show that the proposed model can distinguish the difference of cell survival for cells exposed to different types of particles at different LET.

    Study on Design Method of New Operating Phantom for Gamma Irradiation Processing
    YANG Lei;LIU Yan;ZHOU Yiji;LI Wenge;ZHANG Meiqin
    2020, 54(11):  2282-2288.  DOI: 10.7538/yzk.2020.youxian.0660
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    A new operating phantom was studied for obtaining the true distribution of the radiation field of the gamma irradiation facility and the dose field in the cargo container. An effective and usable simulated radiation field was constructed based on the parameter of the real gamma irradiation facility. The design method for the structure and padding of new operating phantom was proposed. The Monte Carlo method and random filling method (RCS) were adopted to simulate the distribution law of the main factor affecting the dosimetry parameters of the phantom body, such as the dosimeter sleeve material and wall thickness, filler ball size, hollow filler ball size and wall thickness, ball filling method, etc. Under the premise of satisfying the calculation confidence level, the optimization ranges of parameters are as follows: Aluminum tube with wall thickness of 3-5 mm is used as the sleeve, which interferes with the radiation field no more than 4.372 23%; Polypropylene sphere with outer diameter of 1-4 cm, wall thickness of 1.1-11.5 mm, cargo density simulating range of 0.1-0.5 g/cm3, and good material equivalence is selected as the phantom padding, which interferes with the radiation field no more than 9.998 44%; The interference level of the uniform and random filling modes on the radiation field is almost the same, and both are less than 10%. The results show that the current design is feasible and effective, with low investment, more parameters and wider range, and it is suitable for popularization and application.