最新刊期
      卷 43 ， 期 4 ， 2020

        SPECIAL SECTION ON THE 11TH NATIONAL CONFERENCE ON NEW AND RESEARCH REACTORS (PART I)

      • “在第四代核能系统研究中，专家通过SCALE程序模拟分析了氟盐和氯盐快堆的燃料增殖比，为熔盐快堆基盐选择和尺寸设计提供理论依据。”
        ZHOU Xuemei,LI Dongguo,HE Liaoyuan,LIU Guimin
        Vol. 43, Issue 4, Article number: 040001(2020) DOI: 10.11889/j.0253-3219.2020.hjs.43.040001
        摘要：[Background andPurposeThe molten salt fast reactor is one of the advanced nuclear energy systems which has attracted much attention because of its outstanding advantages in fuel proliferation, nuclear waste transmutation and inherent safety. Liquid fluorine salt or chlorine salt is usually used as fuel carrier salt and coolant in molten salt fast reactor, which has high breeding properties.MethodsBased on the dual-fluid molten salt reactor core structure, a comprehensive simulation program SCALE (Standardized Computer Analyses for Licensing Evaluation) were adopted to simulate and calculate the U-Pu fuel breeding ratio of two kinds of fluorine-salt fast reactors and one chlorine-salt fast reactor under the same metal solubility.ResultsThe breeding ratios of different breeding and reflecting layers were simulated, and the changes of neutron energy spectra in fission and breeding zones were analyzed. The breeding ratio of chlorine salt fast reactor increases with the increase of the thickness of the breeding layer and the reflector layer, but the growth rate decreases. When the thickness of the breeding layer of the fast reactor is small, the variation of the breeding layer has little effect on the breeding ratio. When the thickness increases to 60 cm, the variation of the breeding layer thickness has almost no effect on the breeding ratio. The variation of reflector size has no effect on the breeding ratio. The critical state of the reactor core and the neutron energy spectrum in the fission region are not affected by the thickness of the breeder layer and the reflector layer.ConclusionsAt the same temperature and molar ratio, chlorine fast reactor has higher U-Pu fuel breeding ratio than fluorine fast reactor. This provides a theoretical basis for basic salt selection and size design for molten salt fast reactor diffusion.  
        关键词：Chlorine salt fast reactor;Fluorine salt fast reactor;Breeding ratio;Neutron spectrum   
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        发布时间：2024-10-25
      • “在研究堆领域，中国先进研究堆CARR的板状燃料元件表现出良好辐照性能。通过热工水力模拟计算，发现当反应堆功率提升时，堵塞流道内燃料组件温度上升，冷却剂沸腾，功率波动，有助于及早探知和预防堵流事故。”
        DING Li,LUO Beibei,HUA Xiao,NING Bo,QIAO Yaxin
        Vol. 43, Issue 4, Article number: 040002(2020) DOI: 10.11889/j.0253-3219.2020.hjs.43.040002
        摘要：BackgroundPlate-type fuel element exhibits excellent irradiation properties in research reactor while the grid spacing between the fuel element plates is usually very small. Most of the blockage accidents are caused by foreign matters entering the flow channel or fuel swelling due to the narrow coolant flow channel in the core.PurposeThis study aims to analyze the process of blockage accidents, and put forward effective preventive measurements and detection methods.MethodsBased on the investigation of some typical accidents abroad, China advanced research reactor (CARR) was taken as a feature research object, and the thermal hydraulic simulation calculation of CARR core, reactor body, single fuel assembly and cooling circuit outside the reactor was carried out by using the RELAP5/MOD3.2 thermal calculation program. Modeling and analysis of the specific flow channel blockage accident process on CARR was finished.ResultsSimulation results show that the temperature of the fuel assembly in the blocked flow passage will increase when the reactor power is raising, then the coolant will begin to boil, and the power will fluctuate significantly.ConclusionsIt is necessary to strictly control the quality of core operation. The monitoring of neutron flux rate and reactor power, and analyzing fuel temperature are helpful to detect and prevent the further development of the blockage accident.  
        关键词：Plate-type fuel element;Flow channel blockage accident;China Advanced Research Reactor   
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        发布时间：2024-10-25
      • “在核电领域，专家利用Co-SPND探测器的相对热中子灵敏度比，为在线监测热中子灵敏度提供解决方案。”
        ZHANG Hengkai,ZHAO Yuntao,LIU Ao LIU Jiguang,ZHENG Junwei
        Vol. 43, Issue 4, Article number: 040003(2020) DOI: 10.11889/j.0253-3219.2020.hjs.43.040003
        摘要：BackgroundCo-SPND （Co-Self Powered Neutron Detector) are used to measurement the neutron flux in reactor core in an EPR (European Pressurized Reactor) nuclear power unit, but their sensitivity of thermal neutron can not be calibrated in low power level (lower than 25%NP) and in real time operation.PurposeThe study aims to find a new way to monitor the thermal neutron sensitivity for Co-SPND detector in low power level and in normal operation.MethodsThe relative thermal neutron sensitivity rate Cij of two Co-SPND detectors with radial symmetry arrangement was proposed as on-line monitoring parameters of thermal neutron sensitivity. It was the ratio of output currents of these two Co-SPND detectors. This method was applied and verified in the start-up and operation of an EPR nuclear power unit.ResultsThe experimental results show the feasibility of this method in an EPR nuclear power unit.ConclusionsThe method is not only suitable for Co-SPND detectors, but also suitable for other types of SPND detector symmetrically arranged in core to monitor the thermal neutron sensitivity in operation.  
        关键词：Co-SPND detector;Thermal neutron sensitivity;Monitoring in operation;Radial symmetry relationship;The relative thermal neutron sensitivity rate   
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        发布时间：2024-10-25
      • “在核能领域，专家采用MCBMPI程序验证VERA燃耗基准题，分析不同燃耗截面库对计算的影响，为核反应堆设计提供参考。”
        YANG Wankui,YUAN Baoxin,HUANG Huan,WANG Guanbo,ZHANG Songbao
        Vol. 43, Issue 4, Article number: 040004(2020) DOI: 10.11889/j.0253-3219.2020.hjs.43.040004
        摘要：BackgroundMonte Carlo depletion method is based on Monte Carlo neutron transport code and ORIGEN2 point depletion code, it is widely used for neutron transport burnup. However, the continuous energy neutron cross sections of the nuclides in the current evaluation library are less than the cross section used in the depletion calculation. Therefore, the depletion cross sections generated by Monte Carlo transport calculation cannot replace all the cross sections in the basic depletion library.PurposeThis study aims to analyse the influence of different basic depletion libraries on coupled calculation of neutron transport burnup.MethodsThe Monte Carlo transport depletion code MCBMPI (Monte Carlo Burnup code in MPI version) was employed to calculate the new VERA (Virtual Environment for Reactor Applications) depletion benchmark. The influence of different basic burnup cross section on the burnup calculation in transportation was compared and analyzed.ResultsCalculation results show that relative error of keff is less than 8‰ in all the depletion stages, and the mass variation of 235U and 135Xe are less than 4‰ and 5‰ respectively, in the last depletion stage. The variation between thermal and fast neutron spectrum cross section library is small in this benchmark.ConclusionsIt is recommended to select a basic depletion cross section library containing similar neutron spectrum instead of a typical thermal neutron cross section library in real application.  
        关键词：Depletion cross section;Monte Carlo depletion;MCBMPI;VERA benchmark   
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        发布时间：2024-10-25
      • “在核电站辐射源分析领域，专家建立了活化产物源项计算模型，开发了CPAP程序，为核电站设计运行提供解决方案。”
        LYU Weifeng,CHEN Mingliang,HUANG Qianqian,SONG Yu,HONG Hao
        Vol. 43, Issue 4, Article number: 040005(2020) DOI: 10.11889/j.0253-3219.2020.hjs.43.040005
        摘要：BackgroundRadioactive activation products are the main source of radiation for pressurized water reactor nuclear power plant.PurposeThis study aims to create the calculation model and analysis tool for the radioactive activation product of a pressurized water reactor nuclear power plant.MethodFirst of all, pathways of producing the source term of activation products was analyzed, a calculation model for the source term of activation product in the materials outside reactor core was established. Then solutions based on matrix exponential method and Chebyshev rational approximation method (CRAM) were given respectively. Furthermore, a calculation code CPAP for the source term of the out-side core radioactive activation product with a man-machine interface was implemented and benchmarked with existing similar codes worldwide based on neutron activation cases of typical material.ResultsThe benchmark results show that the tolerance of CPAP calculation is within the acceptable range and can be used in engineering applications.ConclusionsWith advantages of good man-machine interface, optional solver and so on, CPAP code makes it very suitable for radioactive activation product evaluation of nuclear power plant during the design, commercial operation and decommissioning stage.  
        关键词：Radioactive activation product;Source term;Pressurized water reactor nuclear power plant   
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        发布时间：2024-10-25
      • “在高温气冷堆领域，专家分析了冷却剂杂质对高温合金性能的影响，为合金腐蚀研究提供新思路。”
        WANG Qiuhao,LI Haoxiang,ZHENG Wei,YIN Huaqiang LI Shengqiang HE Xuedong,MA Tao
        Vol. 43, Issue 4, Article number: 040006(2020) DOI: 10.11889/j.0253-3219.2020.hjs.43.040006
        摘要：BackgroundThe small amount of CO, H2, H2O and CH4 and other impurities existed in the primary coolant of high temperature gas-cooled reactor (HTGR) has important influence on the high temperature properties of high temperature alloys for HTGR. A great number of studies have been carried out on this aspect. Due to the difficulty in controlling the trace impurity content in experimental helium gas during the research process, the distribution of relevant research results is relatively scattered, so appropriate models need to be induced analysed.PurposeThis study aims to review two research methods and their applications for the influence of helium impurity in coolant on the properties of high temperature alloys under the very-high temperature operation of HTGR.MethodsFirst of all, the composition and source of impurities in the primary coolant of HTGR were introduced, and the corrosion mechanism of the alloy in the impure helium of HTGR was analyzed. Then two widely used theoretical models, i.e., the stability diagram for chromium model and the ternary environmental attack (TEA) diagram model, were compared and investigated, and some applications of the two methods were reviewed. Finally, the oxygen partial pressure and carbon activity of helium gas of three HTGR coolants were calculated based on the impurity limit data of high temperature reactor using the stability diagram for chromium model. [Results andConclusionsThe stability diagram for chromium model has been widely used to analyze the effect of helium impurity content on the properties of nickel-chromium alloy materials, and its rationality has been verified. The TEA diagram model can specify the allowable impurity levels and provide the basis for reactor gas impurity control. However, it has not been widely used due to the need for many tests in determining the boundary.  
        关键词：HTGR;High temperature alloy;Coolant;Impurity phase diagram model   
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        发布时间：2024-10-25
      • “在三代核电厂设计中，先导式安全阀的超压保护功能至关重要。专家通过RELAP5程序验证了排放载荷分析方法的合理性，并研究了关键因素，为载荷优化提供建议。”
        WU Dan,WANG Jie,DU Sijia,FANG Hongyu,YU Na
        Vol. 43, Issue 4, Article number: 040007(2020) DOI: 10.11889/j.0253-3219.2020.hjs.43.040007
        摘要：BackgroundPilot operated safety valve is widely used in the design of generation-III nuclear power plant (NPP) to provide overpressure protection function. Compared with safety valves used in generation-II nuclear power plant, pilot operated safety valve has much shorter opening time, hence generates great discharge force. Discharge force analysis is crucial to the design of pressurizer discharge pipe of the 3rd generation advanced nuclear power plant. As a necessary input for stress analysis, discharge force is important for evaluating whether the pipe design and pipe support design is reasonable.PurposeThis study aims to develop and validate a set of dischage force analysis method for 3rd generation NPP.MethodsBased on the previous study of basic discharge force analysis method where the RELAP5 was used to model the thermal-hydraulic process and the force results deduced through post-processing,further validation of the proposed discharge force analysis method was processed by simulating the EPRI PWR Safety and Relief Valve Test in this study. Key influencing factors, such as valve opening time, seal temperature, seal volume, et al, were investigated to give useful suggestions for design.ResultsIt is found that valve opening time, seal temperature, and seal volume has a great influence on the calculation results, and the law of influence is very complicated.ConclusionsIt is proved that the RELAP5 code could be used to model the complex discharge process, and the method developed can be applied to engineering practice.  
        关键词：Discharge force;Valve opening time;Seal temperature;Seal volume   
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        发布时间：2024-10-25
      • “在核反应堆模拟领域，美国阿贡国家实验室和密歇根大学合作开发了PROTEUS-MOC程序，通过TREAT试验堆验证了其高精度和高效率，为核反应堆设计提供新工具。”
        ZHANG Guangchun,LIU Jie,YANG Won Sik
        Vol. 43, Issue 4, Article number: 040008(2020) DOI: 10.11889/j.0253-3219.2020.hjs.43.040008
        摘要：BackgroundThe transient reactor test facility (TREAT) is a test reactor designed to evaluate the performance of nuclear fuels and other structural materials. TREAT is characterized by its complex structure, the use of air as coolant and graphite as moderator, and strong neutron leakage and heterogeneities in both the radial and axial directions. These characteristics challenge most deterministic neutronics simulation codes. The PROTEUS-MOC code, developed at the Argonne National Laboratory in collaboration with the University of Michigan, adopts the 2D method of characteristics (MOC) and the 1D discontinuous Galerkin finite element method (FEM) to treat the radial and axial distributions of angular flux, respectively.PurposeThis study aims to evaluate the accuracy and efficiency of PROTEUS-MOC code in steady-state neutronics calculation of TREAT.MethodsThe 3D TREAT model was radially discretized into unstructured mesh grids and axially discretized into a stack of planes. The macroscopic cross sections were obtained using the Serpent Monte-Carlo code.ResultsPROTEUS-MOC demonstrates satisfactory precisions of eigenvalues by accurately capturing the strong neutron leakage of TREAT. Compared with the eigenvalues obtained by Monte Carlo code Serpent, the relative difference of the eigenvalues obtained by PROTEUS-MOC is merely 0.12%. Additionally, the results showed that the acceleration method TLPCMFD (two-level pCMFD) adopted in PROTEUS-MOC is very effective, speeding up the simulation by a factor of 26.ConclusionsThis work demonstrates that PROTEUS-MOC is capable of producing faithful and accurate results for neutronics calculations of reactor problems with strong neutron leakage and heterogeneities.  
        关键词：Transient reactor test facility;Neutronic simulations;Method of characteristics;Discontinuous finite element method   
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        发布时间：2024-10-25
      • “在核电厂事故后果评价领域，专家采用贝叶斯MCMC方法对大气扩散模型进行不确定性分析，提高了模型预测的可信度，为应急响应提供可靠参考数据。”
        CUI Weijie,CAO Bo,CHEN Yixue
        Vol. 43, Issue 4, Article number: 040009(2020) DOI: 10.11889/j.0253-3219.2020.hjs.43.040009
        摘要：BackgroundAppropriate atmospheric diffusion model is necessary to evaluate the consequences of hypothetical accidents at nuclear power plants. Compared with the traditional uncertainty analysis method, the Bayesian method fully considers the existing observation data. The Markov Chain Monte Carlo (MCMC) method can conveniently combine the Bayesian method and Gaussian plume model.PurposeThis study aims to analyze the parameter uncertainty for improving the credibility of model prediction by using Bayesian MCMC based Gaussian plume model (GPM).MethodsThe Bayesian method and Gaussian plume model were combined to be adopted in the MCMC method. Firstly, the sensitivity of the GPM to several important parameters was analyzed by changing the value of one variable at a time, and then the two parameters with the highest sensitivity were selected using the Bayesian MCMC method for uncertainty analysis.ResultsBy analyzing the MCMC sample sequences, the best fitting of observation data and confidence intervals of simulation results are obtained.ConclusionsThe Bayesian method provides a more reliable confidence interval, which can provide better reference data for emergency response after the accident.  
        关键词：Gaussian plume model;Uncertainty analysis;Bayesian MCMC method;Confidence interval   
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        发布时间：2024-10-25

        ACCELERATOR, RAY TECHNOLOGY AND APPLICATIONS

      • “脉冲激光在集成电路单粒子效应研究中取得新进展，为SEE研究提供新方法。”
        ZHANG Chenguang AN Heng,WANG Yi,LI Cunhui,CAO Zhou,WEN Xuan,XUE Yuxiong
        Vol. 43, Issue 4, Article number: 040201(2020) DOI: 10.11889/j.0253-3219.2020.hjs.43.040201
        摘要：BackgroundPulsed laser haa been widely used in the research of single particle effects (SEE) in integrated circuits and devices. Compared with heavy ion sources, it is easier to obtain spatial and temporal information through pulsed laser-induced SEE.PurposeThis study aims to compare the linear energy transfer (LET) of SEE in silicon devices between single-photon and two-photon.MethodsThe differences of LET were compared by formula derivation and experimental comparison. Three kinds of silicon devices, i.e., differential comparator, static random access memory (SRAM) and bipolar junction transistor, were taken as design under test (DUT) units for both the single-photon absorption (SPA) and two-photon absorption (TPA) experiments.ResultsThe different processes of SPA and TPA induced single particle effect are verified, and comparison results show that there is a quantitative relationship between energy and LET.ConclusionsIt is reasonable to have different equivalence relations between energy and LET. SPA and TPA need different LET values in single particle effect.  
        关键词：Single event effects;Pulsed laser;Single photon;Two-photon;LET   
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        发布时间：2024-10-25
      • “在辐射成像领域，专家基于蒙特卡罗方法建立仿真模型，分析了不同参数对散射的影响，为双投影系统硬件改良提供参考。”
        NI Minzi,WANG Zhentao,CONG Peng,LI Litao
        Vol. 43, Issue 4, Article number: 040202(2020) DOI: 10.11889/j.0253-3219.2020.hjs.43.040202
        摘要：BackgroundIn 60Co dual projection radiation imaging system, the radiation sources are arranged at the bottom and side while their corresponding ionization chamber detectors are set respectively in the gantry. With the synchronous movement of the sources and the gantry, the system can obtain both side-view and upward-view images of the objects simultaneously. However, due to the presence of scatter, the imaging of the two projection planes will interfere with each other.PurposeThis study aims to reduce the effect of scattering between two projection planes and improve the imaging quality of the system by using Monte Carlo method.MethodsFirstly, a simulation model based on Monte Carlo method was established, and its reliability was verified by comparing the simulation data with the experimental data of the scattering distribution at no-load. Secondly, the scattering effects of different projection plane spacings, different shielding materials (tungsten, lead, steel, and aluminum) around the detector with different installation positions, and thickness of the shielding layer on the scatter were calculated based on the model built. Finally, shielding correction effect of 2.5 cm thick lead layer was evaluated.ResultsSimulation results show that when the projection plane spacing is 40 cm and 80 cm, the scattering can be reduced to less than 55% and 40% of that of the 10 cm plane spacing respectively. When projection plane spacing is 40 cm, a lead shield placed between the gantry support and the side-view ionization chambers can achieve best shielding effect at the same thickness compared with steel and aluminum. When the thickness of the lead is 2.5 cm, scattering will be further reduced by 80%.ConclusionsThis study provides a reference and guidance for the hardware improvement and upgrade of the dual-projection system for scattering correction.  
        关键词：60Co dual-projection radiation imaging system;Monte Carlo;Scattering distribution;Hardware correction   
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        发布时间：2024-10-25

        NUCLEAR ELECTRONICS AND INSTRUMENTATION

      • “在静态CT系统领域，研究人员设计了新型X射线管结构，采用超顺排碳纳米管技术，研制出冷阴极X射线管，为脉冲X射线源及静态CT安检系统建设奠定基础。”
        TANG Lihua,ZHANG Guoguang,ZHOU Duanliang,LIU Peng,JIANG Kaili
        Vol. 43, Issue 4, Article number: 040401(2020) DOI: 10.11889/j.0253-3219.2020.hjs.43.040401
        摘要：BackgroundThe cold cathode X-ray tube is a key component of static CT system. Since the discovery of super-aligned arrays of carbon nanotubes, their applications in X-ray tubes have been gradually developed.PurposeThis study aims to design and implement a cold cathode X-ray tube based on super-aligned arrays of carbon nanotubes.MethodsThe cold cathode X-ray tube was developed by designing a new X-ray tube structure and adopting a new X-ray tube preparation process. The acetylene was taken as carbon source gas, the super-aligned arrays of carbon nanotubes were prepared by using the chemical vapor deposition method. Then the cold cathode X-ray tube based on the technology of super aligned carbon nanotubes was developed by taking the nanotubes as electron source, and the performance of the developed X-ray tube was tested. Both the I-V characteristic curve and the X-ray focus size were obtained.ResultsResults shown on the I-V characteristic curve indicates the maximum emission current of this developed cold cathode X-ray tube is 5.0 mA when the gate voltage is 2 373 V. The focus size determined by the pinhole imaging is 0.6 mm×1.6 mm. When the anode high voltage is below 130 kV, the X-ray tube works stably.ConclusionsThis study lays a foundation for the development of pulsed X-ray source and static CT security system.  
        关键词：X-ray tubes;Cold cathodes;Super-aligned arrays;Carbon nanotubes   
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      • “高海拔宇宙线观测站LHAASO的地面簇射粒子阵列KM2A，采用软件触发与数据处理，实现实时分布式算法调度，为高能宇宙线研究提供解决方案。”
        LU Xiaoxu,GU Minhao,ZHU Kejun,LI Fei
        Vol. 43, Issue 4, Article number: 040402(2020) DOI: 10.11889/j.0253-3219.2020.hjs.43.040402
        摘要：BackgroundThe large high altitude air shower observatory (LHAASO) is a major national science and technology infrastructure approved for the 12th Five-Year Plan. It is one of the scientific infrastructures with the highest altitude in China and its core goal is to explore high-energy cosmic ray. As an important component of LHAASO, the ground shower particle array (KM2A) has a huge total area of one square kilometer. Unlike most high-energy physics experiments, the KM2A has a total of about 6 500 channels, and a total input bandwidth of about 546 MB∙s-1. There is no hardware trigger function in KM2A.PurposeThis study aims to design and implement a software for KM2A to complete the trigger and real-time data processing.MethodsData transmission scheme based on standard Ethernet and TCP / IP protocol was employed to receive all packet data including all detector units at the same time. Online processing and calculation of data packets were achieved by using the configured trigger algorithm. Data with physical significance were extracted and packaged to form different events with physical labels, then these events were sent to different receivers according to the configured output rules. By using distributed technology and task division, this software was able to finish real-time data processing through selectable algorithm scheduling.Results & ConclusionsThe software is able to trigger KM2A in real time, reduce the amount of data, and lessen the burden of transmission and storage. It has been put into use in the data center of 4 410 m altitude experiment site, and has been running well so far.  
        关键词：Distributed software;Data acquisition system;Online real-time data process;Software trigger   
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        NUCLEAR PHYSICS, INTERDISCIPLINARY RESEARCH

      • “在石墨去污领域，研究者利用热等离子体技术，通过改变去污时间和电流，有效提高了去污率，为重金属污染治理提供了新思路。”
        SI Mingqiang,ZHANG Jingjing,REN Jinqiu,LU Chunhai
        Vol. 43, Issue 4, Article number: 040501(2020) DOI: 10.11889/j.0253-3219.2020.hjs.43.040501
        摘要：BackgroundDuring the operation of nuclear facilities, a large amount of radioactive waste is produced, most of the radioactive pollutants in decommissioning are surface contamination. Decontamination can effectively remove radioactive contamination and reduce the difficulty of subsequent waste disposal. Conventional physical decontamination and chemical decontamination are time-consuming and expensive whilst little attention has been paid to study of effective but low-cost thermal plasma decontamination.PurposeThis study aims to investigate thermal plasma decontamination of surface contaminated graphite, and explore factors related to decontamination ability of thermal plasma.MethodsGraphite samples contaminated with heavy metal cadmium were taken and prepared as two kinds of samples, i.e., drop-on sample and soaking sample. Decontamination was performed using a thermal plasma device. Decontamination efficiency at different output current and action time were recorded. On the basis of hot plasma dry decontamination, the feasibility and regularity of wet decontamination were also investigated and compared.ResultsExperimental results of hot plasma dry decontamination show that the drip-on sample has a higher decontamination rate and a lower peel thickness compared with soaking sample. When the output current is 60 A and the decontamination time is 100 s, the decontamination rates of drip-on and soaking samples are 92% and 65%, respectively, while the stripping thicknesses are 48 and 117 microns, respectively. Under the same conditions, the wet decontamination has a better decontamination rate than the dry decontamination.ConclusionsThe decontamination ability of hot plasma is basically proportional to the output current and decontamination time, and wet decontamination can greatly improve its decontamination ability, which is a promising decontamination method.  
        关键词：Thermal plasma;Graphite;Surface decontamination;Stripping   
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        发布时间：2024-10-25
      • “在中子孔隙度测井领域，专家验证了可控源测井方法的有效性，为测井仪器放射源替代提供应用意义。”
        YANG Ningning
        Vol. 43, Issue 4, Article number: 040502(2020) DOI: 10.11889/j.0253-3219.2020.hjs.43.040502
        摘要：BackgroundIn neutron porosity logging, different response exists when use D-T source to replace chemical source to measure porosity, which makes the experimental data and interpretation model of traditional chemical source neutron porosity logging difficult to apply.PurposeThis study aims to verify whether the existing density correction method can approximate the response of neutron porosity logging with D-T source instead of chemical source.MethodsFirst of all, response results of water saturated limestone formations with different porosity and different mudstone were obtained by simulation calculation using Monte Carlo N-Particle Transport Code (MCNP), and benchmarked the simulation results with the experimental device measurements. Then these differences among D-T source’s raw data, the D-T source’s corrected data and chemical source’s data were compared and analyzed. Finally, actual logging data were used to validate the effectiveness of the controlled source applicable to neutron porosity logging.ResultsAfter density correction, the results of porosity logging of the controlled source are similar to that of the chemical source, and good correspondence between the two results are found in their actual logging curve.ConclusionsThis study has certain application significance for verifying the validity of control source for neutron porosity logging and the compatible substitution of radioactive sources in future logging tools.  
        关键词：Neutron porosity logging;D-T source;Density correction;MCNP;Logging curve   
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        发布时间：2024-10-25
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