D-T中子源快中子照相准直屏蔽系统设计

减少散射中子对成像质量的影响是快中子照相的关键技术之一。本文采用MCNP程序完成了D-T中子源快中子照相准直屏蔽系统的设计,该设计具有很好的中子散射屏蔽效果,能够将实验室墙壁的散射中子数量降低一个量级,大大减少了成像的对比度失真;同时,能够增加约20%的中子源强,提高了成像速度。     

Measurement of neutron spectra through composed material block bombarded with D-T neutrons

A 2-dimensional composed material assembly made of the iron and hydric block has been established. The neutron spectra from the assembly bombarded with 14-MeV neutrons at neutron generator have been obtained using the proton recoil technique with a stillbene detector. The detector positions were selected at the 60°, 120°, 180° on the surface of the iron spherical shell. The background neutron spectra consisted of background and room return radiation were subtracted with combination of methods of experimental shielding and MCNP calculation. The uncertainty of results was 6.3-7.4%. The experiment results were analyzed and simulated by MCNP code and two data library. The difference is integral neutron flux (background neutron subtracted) of measured results greater than calculations with maximum of 21.2% in the range of 1-16 MeV.     

D-T快中子照相准直屏蔽体设计及中子束特性的模拟研究

设计一个用于氘氚(D-T)快中子照相的准直屏蔽体系统,对D-T中子发生器快中子在准直屏蔽体材料中输运的MCNP模拟研究,给出准直中子束的中子能谱、注量率及均匀性、γ射线能谱和γ射线注量率等重要参数.模拟结果显示,用D-T中子发生器中子源和合理的准直屏蔽体系统可得到快中子照相所需的准直快中子束.     

D-T快中子治疗准直屏蔽体设计及中子特性模拟

设计了一个用于D-T快中子治疗的准直屏蔽体,通过D-T中子在准直屏蔽体中的MCNP模拟,计算了屏蔽体外透射中子和透射光子在水中的吸收剂量,由此评价了准直屏蔽体的屏蔽效果。利用MCNP程序,模拟了准直中子束及中子束中的γ射线在源皮距(SSD)100cm处的能谱,计算了γ射线与中子束在水中吸收剂量的比值,对准直中子束中γ射线的污染水平进行了评价。完成了准直中子束在人体组织等效水箱中输运的MCNP模拟,给出了吸收剂量深度分布、吸收剂量横向分布和吸收剂量等剂量曲线。     

基于氘氚中子源硼中子俘获治疗的中子慢化整形研究

硼中子俘获治疗（Boron Neutron Capture Therapy,BNCT）是一种具有广阔前景的癌症治疗方法。氘氚中子源是未来可供选择的BNCT中子源之一,由于氘氚中子源产生的中子能量为14.1 MeV,不能直接用于BNCT,需要进行束流慢化整形。使用蒙特卡罗模拟程序MCNP5设计了相应的束流整形组件（Beam Shaping Assembly,BSA）,模拟验证了用半径为14 cm的天然铀球做中子倍增层的优越性,计算结果表明：采用50 cm厚的BiF3和10 cm厚的TiF3组合慢化层,17 cm厚的AlF3补充慢化层,0.2 mm厚的Cd热中子吸收层,3.5 cm厚的Pb作为γ屏蔽层,以及10 cm厚的Pb反射层,获得了较为理想的治疗中子束,输出中子束的空气端参数满足国际原子能机构（International Atomic Energy Agency,IAEA）的建议值。     

Tokamak D-T neutron source models for different plasma physics confinement modes

Neutronic studies of European demonstration fusion power plant (DEMO) have been so far based on plasma physics low confinement mode (L-mode). Future tokamaks, nevertheless, may likely use alternative confinement modes such as high or advanced confinement modes (H&A-mode). Based on analytical formulae used in plasma physics, H&A-modes D-T neutron sources formulae are proposed in this paper. For that purpose, a tokamak random neutron source generator, TRANSGEN, has been built generating bidimensional (radial and poloidal) neutron source maps to be used as input for neutronics Monte-Carlo codes (TRIPOLI-4 and MCNP5). The impact of such a source on the neutronic behavior of the European DEMO-2007 Helium-cooled lithium–lead reactor concept has been assessed and compared with previous results obtained using a L-mode neutron source. An A-mode neutron source map from TRANSGEN has been used with the code TRIPOLI-4. Assuming the same fusion power, results show that main reactor global neutronic parameters, e.g. tritium breeding ratio and neutron multiplication factor, evolved slightly when compared to present uncertainties margin. However, local parameters, such as the neutron wall loading (NWL), change significantly compared to L-mode shape: from ?22% to +11% for NWL.     

D-T孔隙度测井数据校正及实验研究

采用氘-氚(D-T)可控源进行中子孔隙度测井时,其响应结果与化学源中子孔隙度测井的响应结果存在差异,使得传统化学源中子孔隙度测井的实验数据和解释模型难以适用。为了验证已有的密度校正方法是否能够用于中子孔隙度测井的结果校正,使D-T源与化学源的响应结果相接近,本文通过模拟获取不同孔隙灰岩含水地层和泥岩中的响应结果,将模拟结果与研制的实验装置测量结果进行基准检测,然后分析D-T源中子孔隙度测井校正前后与化学源中子孔隙度测井对比的响应差异,并利用实际测井数据验证可控源孔隙度测井方法的有效性。结果表明:经过密度校正后,可控源与化学源中子孔隙度的测量结果有较好的相似性,在实际测井曲线上两者也存在较好的对应关系。因此,本研究对于验证可控源中子孔隙度测井方法的有效性和今后测井仪器中放射源的可兼容替代有一定的应用意义。     

The design of “multiple energies from one source” in D-T neutron tube

To obtain multiple monoenergetic neutron sources and realize the on-site calibration of radiation monitoring equipment for nuclear-involved places,the structural characteristics and neutron source features of D-T neutron tube were analyzed;Monte Carlo method was adopted to simulate the effect of interaction between typical materials and different energy neutrons;multilayered shielding materials were combined and optimized to acquire the optimal scheme to shield the neutron sources from the neutron tube.On the base,a tapered alignment filtration construction was designed and Monte Carlo method was employed to simulate the effect of alignment construction.The result showed that the tapered alignment filtration construction can create monoenergetic neutrons including14.1 MeV,0.18 MeV and thermal neutrons and demonstrated good monochrome performance which provides multiple monoenergetic sources for the on-site calibration.     

基于D-T聚变中子源的双功能液态铅锂包层中子学实验

中国双功能锂铅包层(Dual Functional Lithium-Lead,DFLL)是由中国科学院合肥物质科学研究院核能安全技术研究所设计的用于聚变反应堆的液态包层.由于聚变反应堆氚增殖包层的设计高度依赖于中子计算,为验证DFLL包层设计中所使用的核数据库和仿真软件,建立了DFLL包层实验模块,并基于D-T聚变中子...     

On comparison of experimental and calculated neutron energy flux spectra at miniature neutron source reactor (MNSR)

Neutron energy spectrum in Miniature Neutron Source Reactor (MNSR), called Pakistan Research Reactor (PARR-2), is measured employing threshold neutron activation detectors. The calculated neutron spectrum was obtained through modeling the core in detail in three-dimensions employing the transport theory based code WIMS-D/4 and the diffusion theory based code CITATION which was also used as pre-information in the adjustment procedure. A Number of threshold detectors in the form of thin foils are used for spectrum measurements. Gamma activity of irradiated foils was measured with the help of a gamma spectroscopic system consisting of a high efficiency HPGe detector and 8000 channels PC based multi-channel analyzer. STAYNL computer code supplied by International Atomic Energy Agency (IAEA) was used for neutron spectrum adjustment. The group cross-section values and their covariance matrices were derived from the data given in preprocessed cross section libraries in ENDF–6 format of IRDF-90/NMF-G. The comparison between theoretical and experimental work shows good agreement.     

Neutronics analysis of a stacked structure for a subcritical system with LEU solution driven by a D-T neutron source for 99Mo production

The utilization of neutrons markedly affects the medical isotope yield of a subcritical system driven by an external D-T neutron source.The general methods to improve the utilization of neutrons include moderating,multiplying,and reflecting neutrons,which ignores the use of neutrons that backscatter to the source direction.In this study,a stacked structure was formed by assembling the multiplier and the low-enriched uranium solution to enable the full use of neutrons that backscatter to the source direction and further improve the utilization of neutrons.A model based on SuperMC was used to evaluate the neu-tronics and safety behavior of the subcritical system,such as the neutron effective multiplication factor,neutron energy spectrum,medical isotope yield,and heat deposi-tion.Based on the calculation results,when the intensity of the neutron source was 5×1013 n/s,the optimized design with a stacked structure could increase the yield of 99Mo to 182 Ci/day,which is approximately 16％higher than that obtained with a single-layer structure.The inlet H2O coolant velocity of 1.0 m/s and initial temperature of 20℃were also found to be sufficient to prevent boiling of the fuel solution.     

15MeV D-T中子场比释动能测定

本文叙述了用自己研制的组织等效电离室和组装的G-M计数管γ剂量计测定快中子比释动能的方法和测量结果,测量结果总的估计不确定度为±5.2％。并与用美国FWT IC-17A组织等效电离室和FWT GM-2计数管所测的结果进行了比对,两者符合得很好。     

Geant4 simulation of 238U(n,f) reaction induced by D-T neutron source

Knowledge of actinides(n,f) fission process induced by neutron is of importance in the field of nuclear power and nuclear engineering,especially for reactor applications.In this work,fission characteristics of~(238)U(n,f) reaction induced by D-T neutron source were simulated with Geant4 code from multiple perspectives,including the fission production yields,total nubar,kinetic energy distribution,fission neutron spectrum and cumulative γ-ray spectrum of the fission products.The simulation results agree well with the experimental nuclear reaction data(EXFOR) and evaluated nuclear data(ENDF).Mainly,this work was to examine the rationality of the parametric nuclear fission model in Geant4 and to direct our future experimental measurements for the cumulative fission yields of ~(238)U(n,f) reaction.     

MCNP simulation of neutron energy spectra for a TN-32 dry shielded container

This paper presents a numerical analysis of neutron energy spectra for a TN-32 spent fuel dry storage cask using Monte Carlo simulation. The analysis results were compared with experimental measurements to determine the suitability of using such codes for neutron flux calculations in soft-spectrum neutron environments. Complete spent fuel compositions were generated using Scale 4.4a. Variations in source definition and geometry determined that geometric and source simplifications in the computational model have negligible effect on final neutron energy distribution. Variations between experimental and computed spectra at energies above 1 MeV and below 100 keV demonstrated the shortfalls of various detection instruments used to collect the experimental neutron energy spectra data principally because these instruments were calibrated based on high neutron energy spectra. The MCNP calculations were generally in agree with the experimental data, but predicted that the detectors would over-respond to the neutron spectra around a spent fuel dry shielded container. Computed neutron energy spectra were always conservative when compared to experimental spectra.     

Study of the spectra and doses created in the iron-water shielding of a monoenergetic neutron source

Many-group calculations were made for the penetration of neutrons, emitted from monoenergetic sources, through water, iron, andwater-iron systems of finite dimensions; the results of these calculations are presented. The neutron spectra resulting from the passage of such neutrons through water and iron shielding layers were calculated on the twenty-group diffusion-transport approximation, Detailed attention was paid to the high-energy part of the spectrum; certain peculiarities in neutron migration and moderation processes in shielding of the type in question were elucidated. Dose curves D(r) were plotted for neutrons of various energies.By using the superposition principle, the results enable the neutron spectrum to be determined for sources having any arbitrary spectrum.Translated from Atomnaya Énergiya, Vol. 21, No. 1, pp. 27–35, July, 1966.     

多方向加权方法在D-T聚变中子能谱测量中的应用

中子能谱是研究和诊断核反应过程特性最重要的特征量之一。建立了一种新的多方向加权方法用于D-T聚变中子能谱测量:在反冲质子出射方向的多个不同角度上,同时布置探测器,最终的中子能谱由各方向所获取的反冲质子能谱结合相应的权重值确定。Geant4模拟结果显示,多方向加权方法可以提升探测效率和求解结果精度。利用多方向加权方法对高斯分布中子源以及实际的D-T聚变中子能谱进行测量模拟与求解分析,分析结果验证了该方法的可行性和有效性。