Calculation of the wide-angle neutron spectra from the ~9Be(d,xn) reaction in a thick beryllium target

The multi-layer computing model is developed to calculate wide-angle neutron spectra, in the range from0° to 180° with a 5° step, produced by bombarding a thick beryllium target with deuterons. The double-differential cross-sections(DDCSs) for the ~9 Be(d, xn) reaction are calculated using the TALYS-1.8 code. They are in agreement with the experimental data, and are much better than the PHITS-JQMD/GEM results at 15°, 30°, 45° and 60° neutron emission angles for deuteron energy of 10.0 MeV. In the TALYS-1.8 code, neutron contributions from direct reactions(break-up, stripping and knock-out reactions) are controlled by adjustable parameters, which describe the basic characteristics of typical direct reactions and control the relative intensity and the position of the ridgy hillock at the tail of DDCSs. It is found that the typical calculated wide-angle neutron spectra for different neutron emission angles and neutron angular distributions agree quite well with the experimental data for 13.5 MeV deuterons. The multi-layer computing model can reproduce the experimental data reasonably well by optimizing the adjustable parameters in the TALYS-1.8 code. Given the good agreement with the experimental data, the multi-layer computing model could provide better predictions of wide-angle neutron energy spectra, neutron angular distributions and neutron yields for the ~9 Be(d, xn) reaction neutron source.     

冷中子三轴谱仪屏蔽体的蒙特卡罗模拟

冷中子三轴谱仪( CTAS ) 的屏蔽体对于保障工作人员安全、降低散射大厅本底及提高信噪比具有重要的意义。采用蒙特卡罗程序MCNP5 对谱仪各部分屏蔽体进行了计算,并结合Mcstas 程序确定了CTAS 入口处的中子源,大大提高了计算效率。经过模拟计算和优化表明:单色器后端使用厚350mm、密度4.6 g/cm3 的重混凝土,衔接屏蔽体使用厚300 mm、密度3.6 g/cm3的重混凝土,生物屏蔽采用厚150 mm、密度3.6 g/cm3 的重混凝土可保证屏蔽体外表面的剂量率满足散射大厅的剂量要求。The shielding of Cold neutron Triple-Axis Spectrometer( CTAS ) is important for radiation safety of workers, and reduce the background of scattering hall as well as enhancing the ratio of signal-to-noise. In this study,Monte-Carlo simulation was performed to conduct the calculation on the shielding of CTAS. To increase the calculation efficiency, neutron source was obtained by using Mcstas code. The results indicate that, in the case of heavy concrete ( density 4.6 g/cm3 ) with thickness of 350 mm for the shielding behind the monochromater, and heavy concrete ( density 3.6 g/cm3 ) with thickness of 300 mm for the other monochromater shielding, as well as the heavy concrete ( density 3.6 g/cm3 ) with thickness of 150 mm for biological shielding, the dose rate outside shielding may meet the requirement of national standard of China.     

新一代堆芯多群常数库制作系统的研制

随着新一代组件计算程序计算精度要求的不断提高,组件计算程序对配套多群常数库提出了更高的要求,例如:要求能群结构更加精细、共振参数更加多样、燃耗数据更加精确等。为满足新一代组件计算程序的这一系列需求研发了一套制作多群常数库的系统NPLC-3。NPLC-3系统以NJOY程序为中心包含了输入参数库、驱动程序、主库制作程序、并群、并库、删除、添加、进制转换、数据自动检验等功能模块,针对各功能模块采用不同方法初步验证了系统的正确性。相对传统的数据制作方法,NPLC-3系统在燃耗链设计、燃耗数据计算以及共振数据制作等方面增加新的特性,而且系统采用独立的I/O接口、全自动读取参数库并生成输入卡,相对过去制作数据库的手段有了很大的改进。With the unceasing enhancement of the calculation accuracy requirement on the new generation of lattice code, higher requirements on multi-group constants library were put forward. For example, fine energy group structures are required, more types of data are need for the resonance processing, more accuracy burn-up data is required. To meet this series of demand, a multi-group constants production system NPLC-3 was developed. The NPLC-3 system which mainly based on the NJOY program contains an input parameter database and a series of functional codes such as driver code, main library production code, energy group collapsing code, work library production code and so on. Recently, different methods were adopted to validate these codes respectively according to their functions. Compared to the traditional methods in multi-group constants production, NPLC-3 system adopts several new methods in the design of burn up chain, calculation of burn up data and resonance parameters. What's more, the NPLC-3 system has an independent I/O interface, and can fully automatic generate input cards from the input parameters database. Relative to the past means of production library, NPLC-3 system has great improvement.     

Radiation emission from the ＂Rhodotron-TT200＂ cavity

In this paper, the X-rays emitted from the Rhodotron-TT200 cavity have been studied in depth. We found that the Bremsstrahlung interaction is the only contribution of X-ray generation important to safety. The X-ray dose rate in the Rhodotron vault is calculated for normal conditions based on MCNP4C results. The presented calculation shows good agreement with the experimental measurements, which consequently confirms the reliability of the calculation for use in shielding design and other safety aspects.     

HIAF BRing及高能外靶实验终端的辐射屏蔽设计

强流重离子加速器（HIAF）是中国科学院近代物理研究所自主研制的一台高能强流重离子加速器,它可以实现p到U的全离子加速。为了保证HIAF运行时的辐射安全,针对该装置的增强器（BRing）及高能外靶实验终端,利用蒙特卡洛程序FLUKA及外推法计算得到了加速p,C及U三种离子时所需的辐射屏蔽。结果表明,加速质子时所需屏蔽厚度最大,并以此为依据给出了全地下结构的屏蔽设计。在此基础上,提出了一种估算高能质子/重离子加速器束流均匀损失时横向屏蔽厚度的方法。结果显示,估算结果与FLUKA计算结果符合较好,验证了该方法的有效性和准确性。High Intensity heavy-ion Accelerator Facility (HIAF) is designed by the Institute of Modern Physics, Chinese Academy of Sciences, which can accelerate particles from proton up to uranium. To guarantee the radiation safety of HIAF during operation, the FLUKA code and extrapolation method were adopted to calculate the shielding thickness. The calculations were based on proton, carbon and uranium particles when losing on the Booster Ring (BRing) and the high-energy experimental terminal. The results indicate that the shielding thickness required for accelerating protons was the largest. Basing on the results, a method for estimating the lateral shielding of a high-energy proton/heavy-ion accelerator was proposed. A good agreement shows between the estimated results and the FLUKA calculated results, the validity and accuracy of the method were verified.     

Monte Carlo simulation for bremsstrahlung and photoneutron yields in high-energy x-ray radiography

This paper reports on the results of calculations using a Monte Carlo code (MCNP5) to study the properties of photons, electrons and photoneutrons obtained in the converted target and their transportations in x-ray radiography. A comparison between measurements and calculations for bremsstrahlung and photoneutrons is presented. The radiographic rule and the effect of the collimator on the image are studied with the experimental model. The results provide exact parameters for the optimal design of radiographic layout and shielding systems.     

γ辐射多层组合屏蔽的蒙特卡罗方法模拟及其论证

应用蒙特卡罗（MC）方法通用软件(MCNP4B程序)建立多层组合模型, 模拟计算了γ辐射非均匀屏蔽的问题, 比较研究各种组合方法以确定最佳的方案。 同时对模拟结果与理论计算、 实验测量三者进行综合分析, 说明了三者结论的一致性, 也说明了用MC方法来模拟非均匀屏蔽的可行性和多层组合屏蔽设计的实际意义。 In this paper, the multi layer model was established to calculate γ radiation non uniform shielding problem, and various combinations of methods were investigated to determine the optimal option. At the same time, the comprehensive analysis of the simulation results, the theoretical calculation and the experimental measurements show that the consistency among them. Practical significance is also shown that the application of Monte Carlo method to simulate the non uniform shielding feasibility and multi layer shielding design.     

Study on the radiation problem caused by electron beam loss in accelerator tubes

The beam dynamic code PARMELA was used to simulate the transportation process of accelerating electrons in S-band SW linacs with different energies of 2.5, 6 and 20MeV. The results indicated that in the ideal condition, the percentage of electron beam loss was 50% in accelerator tubes. Also we calculated the spectrum, the location and angular distribution of the lost electrons. Calculation performed by Monte Carlo code MCNP demonstrated that the radiation distribution of lost electrons was nearly uniform along the tube axis, the angular distributions of the radiation dose rates of the three tubes were similar, and the highest leaking dose was at the angle of 160° with respect to the axis. The lower the energy of the accelerator, the higher the radiation relative leakage. For the 2.5MeV accelerator, the maximum dose rate reached 5% of the main dose and the one on the head of the electron gun was 1%, both of which did not meet the eligible protection requirement for accelerators. We adopted different shielding designs for different accelerators. The simulated result showed that the shielded radiation leaking dose rates fulfilled the requirement.     

WIMS多群截面库更新时NJOY输入参数研究

以最新的微观评价库CENDL-2.1为基础,用先进的群常数制作程序系统——NJOY来制作新的WIMS69群截面库,研究不同的 NJOY输入参数对 WIMS程序所计算的积分量的影响,并给出了详细的参数研究结果,同时还分析和讨论了计算结果与基准实验结果的比对.WIMS multi group constant library is the associated working library of WIMS/D4 lattice code, and it was created by using rather old and obsolete data based on ENDF/B3 (1972). Recently, the new evaluated data files such as ENDF/B 6.5, JEF 2.2, CENDL 2.1 and JENDL 3.2 were released. It s necessary to update the old library by the new evaluated data. The parameter study is performed to investigate the sensitivity of the integral parameters calculated with WIMS/D4 on the selection...     

中子、γ射线在稀土-高分子材料中的输运(英文)

为研究新型复合屏蔽材料的最佳厚度与各种成分最佳配比, 用MCNP计算了中子、 γ射线在稀土 高分子与重金属复合材料中的通量。 对中子、 γ射线在屏蔽体中变化规律进行了深入探索, 同传统复合屏蔽材料的屏蔽性能进行了对比。 结果表明, 中子和γ射线通过屏蔽体时, 其强度遵循指数衰减规律。 新型屏蔽材料对中子的屏蔽效果均优于铅硼聚乙烯, 对γ射线的屏蔽效果均劣于W Ni合金, 且并非稀土含量越高, 材料对中子辐射屏蔽能力越强。 A series of shielding analyses have been performed to estimate the material composition and optimum thickness required for a new radiation shield with various rare earth doped polymer and heavy metal mixtures. The neutron and γ photon fluxes have been calculated by Monte Carlo N Particle(MCNP) transport code. The results indicate that the relative fluxes of γ photon and neutron in both traditional and new composite materials follow an exponential decay rule with the distance of penetration. It can be seen that the composite material consisting of rare earth doped polymer and heavy metal has stronger neutron shielding performance than lead boron polyethylene, but weaker γ shielding effectiveness than W Ni alloy. It is also found that materials with more components of rare earth elements don’t always provide better neutron shielding performance.     

蒙特卡罗输运模拟软件JMCT的深穿透屏蔽计算

核设施辐射屏蔽计算,由于其大规模计算及深穿透等特性,一直是蒙特卡罗方法工程应用的难点之一。采用我国自主研发的三维中子-光子蒙特卡罗粒子输运模拟软件JMCT,结合可视化建模工具JLAMT,对OECD国际基准例题Winfrith Iron/Water Benchmark Experiment(ASPIS)两例实验装置进行建模与计算分析, 并将计算结果与实验值及MCNP计算值进行对比。结果表明,JMCT计算值与MCNP计算值符合较好,其中Winfrith Iron Benchmark Experiment(ASPIS)最大偏差不超过7%,平均偏差1.3%;Winfrith Water Benchmark Experiment(ASPIS)最大偏差小于20%,平均偏差小于10%,证明了JMCT在屏蔽计算以及深穿透问题的可靠性与工程应用性。     

HL-2A中子相机屏蔽系统模拟

利用蒙特卡罗模拟程序,建立了HL-2A中子相机蒙特卡罗粒子输运(MCNP)物理模型,对D-D聚变中子和γ射线的屏蔽进行了模拟计算。对石蜡碳酸锂混合物、聚乙烯、铅和316L不锈钢4种常用中子慢化吸,收剂组成的屏蔽层材料的屏蔽效果进行了对比。计算结果表明,石蜡碳酸锂混合物和铅组合是中子相机的最佳屏蔽层材料,其中石蜡碳酸锂混合物用于慢化吸收中子,铅用于屏蔽中子和γ射线。此外,利用MCNP模拟计算得到了屏蔽中子和γ射线所需的屏蔽厚度,以及准直管的中子散射率。     

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The physical model of the neutron camera Monte-Carlo partical transport (MCNP) for HL-2A was established by using Monte-Carlo simulation code. The shielding of D-D fusion neutrons and gamma rays was simulated. The shielding effects were compared for four common shielding materials, including mixture of paraffin lithium carbonate, polyethylene, lead, 316L stainless steel. Calculation results show that mixture of paraffin lithium carbonate and lead are the best shielding materials for neutron camera, among them the mixture of paraffin lithium carbonate is used for slowing-down and absorpting neutrons, while lead is used for blocking neutrons and gamma rays. In addition, both the required thickness of shielding material for neutron and gamma ray and the neutron scattering rate of collimator tube have been obtained by using MCNP simulation.     

基于VENUS-III国际基准模型的JMCT程序验证

基于自主研制的三维中子-光子耦合输运蒙特卡罗通用程序JMCT（J Monte Carlo Transport Code）,采用连续点截面,对国际基准屏蔽VENUS-III模型开展精细建模和中子输运临界及屏蔽计算.临界计算得到系统k_eff、重要区域的通量及能谱.结果表明,JMCT和MCNP程序的重要区域体通量计数吻合较好,偏差均在1%以内.深穿透屏蔽计算采用外源模式,点探测器计数,JMCT计算值与实验测量值偏差在15%以内,满足屏蔽设计对误差的要求.初步验证了JMCT程序临界及屏蔽计算的可用性.     

基于MCNP计算结果的快速物理插值方法（英）

在核辐射科学研究中经常使用MCNP蒙卡计算程序,但其计算结果只是给定位置的物理量。为了快速获得计算点附近的感兴趣位置处的值,本研究建立了一种3点插值计算方法。首先分析了粒子在材料中输运的物理规律,给出了满足3点插值的空间范围,建立了插值计算模型。然后利用距离平方反比权重法,结合粒子衰减规律,推导了插值计算公式。最后设计了计算机程序,并用X射线屏蔽后注量的MCNP计算结果进行验证。通过和直接采用的距离平方反比权重法和克里金插值法对比,考虑物理作用机制的该计算方法精度更高,且误差在10%以内。     

Determining the gamma-ray parameters for BaO–ZnO–B2O3 glasses using MCNP5 code: a comparison study

Rapid technological advancement has multiplied people’s exposure to ionizing radiations greatly. Widespread applications of radiation in different fields (such as agriculture, radiation therapy and scientific research fields) require that humans be protected against unnecessary exposure. In this study, mass attenuation coefficient (μ_m), half-value layer, mean-free path, effective atomic number (Z_eff) and exposure buildup factor have been calculated for xBaO–20ZnO–(80???x)B₂O₃ (x?=?5, 10, 15, 20 and 25?mol%) glass systems. The mass attenuation coefficients of the selected glasses were calculated using simulation method of MCNP5 code. The simulation results have been compared with the experimental data and Xcom at the energies 223.02, 252.98, 287.28, 340.83, 398.97, 481.59, 562.68 and 662.00?keV. The agreement amounts of the mass attenuation coefficient values are from 0.2% to 2.8% and from 0.2% to 6.98% for MCNP5 and Xcom relative to experimental results, while the Monte Carlo program values are higher than that obtained by experimental data, using Xcom and MCNP5 code. The glass sample having the highest value of BaO content show high radiation shielding properties. It indicates that the MCNP5 code can be used for estimation of radiation interaction parameters where experimental results are not available.     

Monte Carlo simulation for the estimation of iron in human whole blood and comparison with experimental data

Monte Carlo N-particle (MCNP) code has been used to simulate the transport of gamma photon rays of different energies (22, 31, 59.5 and 81 keV) to estimate the iron content in solutions. In this study, MCNP simulation results are compared with experiment and XCOM theoretical data. The simulation shows that the obtained results are in good agreement with experimental data, and better than the theoretical XCOM values. The study indicates that MCNP simulation is an excellent tool to estimate the iron concentration in the blood samples. The MCNP code can also be utilized to estimate other trace elements in the blood samples.     

Comparison study of photon attenuation characteristics of Lead-Boron Polyethylene by MCNP code,XCOM and experimental data

The linear attenuation coefficient, mass attenuation coefficient and mean free path of various Lead-Boron Polyethylene (PbBPE) samples which can be used as the photon shielding materials in marine reactor have been simulated using the Monte Carlo N-Particle (MCNP)-5 code. The MCNP simulation results are in good agreement with the XCOM values and the reported experimental data for source Cesium-137 and Cobalt-60. Thus, this method based on MCNP can be used to simulate the photon attenuation characteristics of various types of PbBPE materials.     

基于蒙特卡罗的小尺度参考辐射装置屏蔽研究

针对移动式小尺度参考辐射（MRR）装置（移动式校准装置）,在进行射线辐射剂量测量的仪器仪表标定或刻度时,应满足其辐射屏蔽安全限值5 Sv/h的屏蔽技术要求,采用蒙特卡罗输运程序MCNP,开展了移动式小尺度参考辐射装置表面剂量场屏蔽的模拟计算和研究分析工作。研究结果表明,通过MCNP模拟的屏蔽设计方法可以详尽反映MRR装置各个表面的剂量分布特征和规律,实现移动式小尺度参考辐射装置屏蔽设计,采用的铅钢材料复合屏蔽方案能够保证装置硬度且显著地减轻屏蔽体的重量,最终获取的优化MRR屏蔽箱体重量约为271.9 kg。