Design calculation of an epithermal neutronic beam for BNCT at the Syrian MNSR using the MCNP4C code

This article describes the design calculation of an epithermal neutronic beam for the boron neutron capture therapy at the Syrian MNSR by using the MCNP4C code and ENDF/B-V cross-section library. To produce a high flux of epithermal neutrons at the beam exit, the moderator/filter from Al, Cd, Fluental and Bi was used with Pb as reflector for neutrons along the beam. In addition, the Bi lined collimator with Li₂CO₃-PE and Pb at the end. The calculated beam parameters under 30.0 kW of reactor power at the beam exit are Ф_epi = 2.83 × 10⁸ n/cm² s, D_f/Ф_epi = 7.98 × 10⁻¹¹ cGy cm²/n, D_γ/Ф_epi = 1.70 × 10⁻¹¹ cGy cm²/n, Φ_epi/Φ_the = 0.05 and J_n⁺/Ф_n = 0.77. As well as, the calculated values of the advantage depth and advantage ratio are 7.51 cm and 3.49, respectively. If such beam was built into the Syrian MNSR the scientific applications of the reactor would increase.     

9Be(d,n)加速器中子源中子照相的研究

加速器中子源比反应堆中子源更具灵活性,北京大学正在发展基于RFQ加速器的小型中子照相装置.为了更好地设计和优化此装置,实现高品质的中子照相,我们在北京大学4.5 MV静电加速器上建立了中子照相实验平台,包括科学级制冷、高灵敏度、低噪声的CCD数字成像系统,模拟基于厚铍靶9Be(d,n)反应RFQ中子源的条件,并利用此系统开展中子成像技术的研究.实验在像平面热中子注量率为5×103 cm-2·s-1或快中子注量率为3.7×104 cm-2·s-1的情况下获得了一定质量的热中子及快中子照片.当利用RFQ直线加速器强中子源时将可获得更高质量的图片,从而可以满足大多数的应用需要.     

热中子照相的MCNP数值模拟初步研究

中子照相的数值模拟在中子照相技术研究中有着重要的作用。介绍了MCNP程序在数值模拟中子照相中的应用,给出了样品的模拟图像结果,并利用MCNP模拟分析了散射中子对中子照相成像结果的影响,对以后的中子照相实验具有指导意义。     

Monte Carlo simulation of core physics parameters of the Syrian MNSR reactor

A 3-D neutronic model for the Syrian Miniature Neutron Source Reactor (MNSR) was developed earlier to conduct the reactor neutronic analysis using the MCNP-4C code. The continuous energy neutron cross sections were evaluated from the ENDF/B-VI library. This model is used in this paper to calculate the following reactor core physics parameters: the clean cold core excess reactivity, calibration of the control rod and calculation its shut down margin, calibration of the top beryllium shim plate reflector, the axial neutron flux distributions in the inner and outer irradiation positions and calculations of the prompt neutron life time (_lp$l_p$) and the effective delayed neutron fraction (_βeff${\beta }_{\mathit{eff}}$). Good agreements are noticed between the calculated and the measured results. These agreements indicate that the established model is an accurate representation of Syrian MNSR core and will be used for other calculations in the future.     

Time dependent burn-up and fission products inventory calculations in the discharged fuel of the Syrian MNSR

The Syrian Miniature Neutron Source Reactor (MNSR), a 30 kW, 89.8% HEU fueled (U-Al), went critical in March, 1996. By operating the reactor at nominal power for 2.5 h/day, the estimated core life is 10 years. This paper presents the results of fuel burn-up and depletion analysis of the MNSR fuel lattice using the ORIGEN 2 code. A one-group cross-section data base for the ORIGEN 2 computer code was developed for the Syrian MNSR research reactor. The ORIGEN 2 predicted burn-up dependent actinide compositions of MNSR spent fuel using the newly developed data base show a good agreement with the published results in the literature. In addition, the burn-up characteristics of MNSR spent fuel was analyzed with the new data base. Finally, to study the effect of burn-up on the reactivity, the microscopic cross-sections of the fission products calculated by the WlMS code (using the number densities of fission products generated by the ORIGEN 2 code as a function of burn-up time), were used as an input for the CITATION code calculations. The results contained in this paper could be used in performing criticality safety analysis and shielding calculations for the design of a spent fuel storage cask for the MNSR core.     

热中子照相成像系统的研制

成像系统是中子照相装置的关键组件之一.利用Gd和In金属研制了胶片静态照相转换屏,采用6LiF ZnS为转换材料,增强型CCD和制冷型CCD相结合,研制了CCD在线中子成像系统.     

Design of a mobile neutron radiography installation based on a compact sealed tube neutron generator

A series of optimum conditions are taken into account in the construction of neutron radiography(NR) installation based on a sealed tube neutron generator capable of gnerating 10^10 n/s with 14MeV.The characteristics of NNU screens,a kind of self-made ^6LiF.ZnS(Ag)scintillation intensifying screen are presented.Finally,some neutron radiographies taken by this NR installation and NNU screens are given.     

Availability of MCNP & MATLAB for reconstructing the water-vapor two-phase flow pattern in neutron radiography

The China Advanced Research Reactor (CARR) is scheduled to be operated in the autumn of 2008. In this paper, we report preparations for installing the neutron radiography instrument (NRI) and for utilizing it efficiently. The 2-D relative neutron intensity profiles for the water-vapor two-phase flow inside the tube were obtained using the MCNP code without influence of γ-ray and electronic-noise. The MCNP simulation of the 2-D neutron intensity profile for the water-vapor two-phase flow was demonstrated. The simulated 2-D neutron intensity profiles could be used as the benchmark data base by calibrating part of the data measured by the CARR-NRI. The 3-D objective images allow us to understand the flow pattern more clearly and it is reconstructed using the MATLAB through the threshold transformation techniques. And thus it is concluded that the MCNP code and the MATLAB are very useful for constructing the benchmark data base for the investigation of the water-vapor two-phase flow using the CARR-NRI.     

Calculation of fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor using the GETERA code

Calculations of the fuel burnup and radionuclide inventory in the Syrian miniature neutron source reactor (MNSR) after 10 years (the reactor core expected life) of the reactor operation time are presented in this paper using the GETERA code. The code is used to calculate the fuel group constants and the infinite multiplication factor versus the reactor operating time for 10, 20, and 30 kW operating power levels. The amounts of uranium burntup and plutonium produced in the reactor core, the concentrations and radionuclides of the most important fission products and actinide radionuclides accumulated in the reactor core, and the total radioactivity of the reactor core were calculated using the GETERA code as well. It is found that the GETERA code is better than the WIMSD4 code for the fuel burnup calculation in the MNSR reactor since it is newer, has a bigger library of isotopes, and is more accurate.     

ZnS、聚丙烯混合快中子荧光屏14MeV中子照射时的性能初步理论分析

分析ZnS、聚丙烯混合压制的快中子荧光屏发光机理,建立了转换屏发光的数学模型,计算获取了入射中子能量为14MeV时,荧光屏厚度与输出光强的变化曲线、荧光屏的点扩散函数、荧光屏材料配比与输出光强的变化曲线,取得了一些有意义的结果,为14MeV快中子荧光屏的参数优化提供了理论基础.     

SPRR-300反应堆中子照相装置准直系统的研制

准直器是中子照相装置的关键组件之一．利用石墨为慢化剂．铋为γ过滤器。采用光阑直径、准直器长度、石墨塞长度可调以实现准直比和镉比的大幅度调节．钢结构水箱快门完成了SPRR—300准直系统设计。实验结果显示设计是合理的．成像质量达ASTM-86达二级以上．能满足不同成像技术和不同样品的中子照相需求。     

单球多计数器的中子能量响应计算

根据球体内随深度变化时中子的慢化程度有所差异．按两两垂直的方法把三个位置灵敏正比计数器安装在一个慢化球体内。用MCNP4A程序计算了6种慢化球体和6种照射方向的能量响应,同时对球半径方向两种分区方法的计算结果进行分析和比较。     

Design of a permanent epithermal neutron irradiation site in the Syrian MNSR

A permanent epithermal neutron irradiation site was designed in the Syrian Miniature Neutron Source Reactor (MNSR) by using cadmium as a thermal neutron shielding material. This site was designed by Cd-shielding the internal surface of the outer aluminum tube of the FOIS (First Outer Irradiation Site) in the MNSR. The MCNP-4C calculations showed that, to have a permanent epithermal neutron irradiation site for the ENAA using the cadmium, it is necessary to add the top beryllium shims of the reactor to compensate for the reactivity losses due to the neutrons absorption in the cylindrical cadmium shell. The activation detectors were used to measure the thermal and epithermal neutron fluxes in the FOIS. Distribution of the thermal neutron flux along the vertical direction of the outer irradiation capsule used in the FOIS has been found using MCNP-4C code, and experimentally by irradiating five copper wires. Good agreements were obtained between the calculated and the measured results.     

Si-PIN探测器中子直照灵敏度研究

介绍了利用K600中子发生器进行Si-PIN探测器灵敏度标定的实验方法,并在实验中测出了Si-PIN探测器对14MeV中子的直照灵敏度。同时,利用MCNP模拟程序对Si-PIN探测器不同能量的中子直照灵敏度进行了理论计算,实验灵敏度处理结果和理论计算值较为一致。     

Thermoluminescence measurements of neutron streaming through JET Torus Hall ducts

Thermoluminescence detectors (TLD) were used for dose measurements at JET. Several hundreds of LiF detectors of various types, standard LiF:Mg,Ti and highly sensitive LiF:Mg,Cu,P were produced. LiF detectors consisting of natural lithium are sensitive to slow neutrons, their response to neutrons being enhanced by ⁶Li-enriched lithium or suppressed by using lithium consisting entirely of ⁷Li. Pairs of ⁶LiF/⁷LiF detectors allow distinguishing between neutron/non-neutron components of a radiation field. For detection of neutrons of higher energy, polyethylene (PE-300) moderators were used. TLDs, located in the centre of cylindrical moderators, were installed at eleven positions in the JET hall and the hall labyrinth in July 2012, and exposure took place during the last two weeks of the experimental campaign. Measurements of the gamma dose were obtained for all positions over a range of about five orders of magnitude variation. As the TLDs were also calibrated in a thermal neutron field, the neutron fluence at the experimental position could be derived. The experimental results are compared with calculations using the MCNP code. The results confirm that the TLD technology can be usefully applied to measurements of neutron streaming through JET Torus Hall ducts.     

在特定实验条件下的散射中子本底研究

研究了d-T中子源与探测器距离较近时,扣除实验大厅散射中子本底的方法。实验上采用屏蔽法,用了铀裂变电离室。用MCNP/4A程序和FENDL2库数据计算了实验大厅散射中子本底曲线。采用实验和计算相结合的方法扣除了在特定实验条件下的散射中子本底,方法是可行的。     

Fluence and dose measurements for an accelerator neutron beam

The 3 MV Van de Graaff accelerator at McMaster University accelerator laboratory is extended to a neutron irradiation facility for low-dose bystander effects research. A long counter and an Anderson-Braun type neutron monitor have been used as monitors for the determination of the total fluence. Activation foils were used to determine the thermal neutron fluence rate (around 10⁶ neutrons s⁻¹). Meanwhile, the interactions of neutrons with the monitors have been simulated using a Monte Carlo N Particle (MCNP) code. Bystander effects, i.e. damage occurring in cells that were not traversed by radiation but were in the same radiation environment, have been well observed following both alpha and gamma irradiation of many cell lines. Since neutron radiation involves mixed field (including gamma and neutron radiations), we need to differentiate the doses for the bystander effects from the two radiations. A tissue equivalent proportional counter (TEPC) filled with propane based tissue equivalent gas simulating a 2 μm diameter tissue sphere has been investigated to estimate the neutron and gamma absorbed doses. A photon dose contamination of the neutron beam is less than 3%. The axial dose distribution follows the inverse square law and lateral and vertical dose distributions are relatively uniform over the irradiation area required by the biological study.     

Efficiency-determined method for thermal neutron detection with inorganic scintillator

Because of 3He shortage, sintillator is a promising alternative choice for neutron detection in the field of thermal neutron scattering and imaging. Also, the neutron detection efficiency is difficult to be determined. In this paper, the efficiency for thermal neutron detection is presented by inorganic scintillator using probability principles, supposed that the material of scintillator is uniform in element distribution, and that attenuation length of scintillation light is longer than that of its thickness in the scintillator. The efficiencies for two pieces of lithium glass are determined by this method, indicating the method is useful for determining efficiency of thermal neutron detections.     

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.