强流氘氚中子源用TiD2靶膜制备技术研究

基片镀膜是氘/氚靶制备过程的重要工序,靶膜的性能直接影响充氘及中子实验。本文对去除表面污渍和氧化层后的基片采用磁控溅射进行镀膜,研制性能优良的强流氘氚中子源用靶膜。采用扫描电镜观察膜层表面外观形貌,根据称重法用电子天秤测量理论膜厚,使用划痕仪分析膜层结合力,并通过电子探针分析膜层的杂质元素含量来表征靶膜的性能。结果表明,磁控溅射镀膜后膜层颗粒度细小、分布均匀,同时膜层表面杂质小于6.0%。镀膜后基片的活化充氘实验表明,氘/钛(原子比)最高可达1.98,满足中子产额实验要求,可进行后续中子实验。     

ADS熔盐散裂靶中子学性能研究

与传统加速器驱动次临界系统（ADS）采用金属靶作为散裂中子靶的设计不同,加速器驱动次临界熔盐堆（AD-MSRs）采用靶堆一体的设计,直接使用燃料熔盐作为散裂中子靶。由于熔盐靶的中子学性能直接影响AD MSRs的能量放大系数、核废物的嬗变和核燃料增殖的效率,所以本研究基于MCNPX程序,详细计算了高能质子轰击氟盐和氯盐两种熔盐靶产生的散裂中子产额、散裂中子能谱、能量沉积分布以及散裂产物等中子学性能,并与液态Pb和铅铋共熔体（LBE）两种液态金属靶进行了对比。计算结果表明,熔盐靶在散裂中子产额上与液态金属靶有一定的差距,但熔盐靶内能量沉积分布的梯度较小,更有利于靶区的热量导出。与液态Pb和LBE靶相比,熔盐靶的散裂产物中包含更多的气体以及高质量数的α发射体核素。     

离子源及厚靶参数对氘氚反应中子源中子产额的影响

用厚靶氘氚(D-T)反应中子产额的计算方法模拟计算了入射氘离子能量为120 keV时D-T中子源的中子产额。研究了氘离子源产生的束流中单原子氘离子(D+)及双原子氘离子(D2+)比例对中子产额的影响。结果表明,提高D+比例,同时降低D2+比例将有效提高中子产额。另外还研究了不同靶膜材料及组分引起的中子产额变化。表明中子产额与靶膜中氚的含量成正比,与靶膜元素的原子质量成反比。同时分析讨论了离子源品质及靶参数对中子源整体性能的影响,得出离子源束流品质的提高对中子源整体的设计至关重要。最后,模拟计算了靶膜表面有氧化层情况下中子产额的变化,并与实验结果作了对比。在此基础上提出了一种新的靶设计方案,并对其物理可行性进行了研究。     

铅散裂中子靶物理特性研究

利用蒙卡程序DCM/CEM对ADS标准散裂中子靶进行了计算。计算了长0.6 m,直径0.2 m的圆柱形208Pb靶,在0.1～1.6 GeV的质子轰击下,Pb靶发生散裂反应产生的中子产额及表面的中子注量、能谱分布以及靶内能量沉积分布,解释了以前的实验结果。计算结果与文献数据、实验数据进行了比较,符合良好,对进一步进行ADS堆芯设计具有较好的理论指导意义。     

加速器驱动系统的靶物理计算分析

通过分析ADS对靶的要求,选LBE（铅铋共熔体）作为ADS靶材料,采用MCNPX程序计算和分析不同半径的质子束流入射不同半径和高度的靶时中子产生数、靶的中子泄漏能谱、靶不同位置的泄漏中子数以及整个靶的泄漏中子总数。选定靶参数,然后通过SSW卡和SSR卡连接MCNPX和MCNP程序,模拟计算1个质子与靶反应直到被燃料利用的整个过程,并提出质子效率的概念,采用欧洲MUSE-4燃料,对靶在堆芯中的位置和靶在堆芯中的半径大小对质子效率的影响进行了计算研究,给出了最大化质子效率时靶在堆芯中的位置和靶半径大小的方案。     

^18O离子轰击厚靶的中子剂量学参数

用阈探测器活化法测量了每个核子动能（简称单核能）为５０ＭｅＶ＾１８Ｏ离子轰击厚靶（Ｂｅ、Ｃｕ、Ａｕ）时出射中子的能量分布、注量率分布和中子角分布，得到了＾１８Ｏ离子的中子产额、前向中子发射率和中子剂量当量率分布的数据。并与国外报道的单核能在２０ＭｅＶ以下的低能重离子轰击厚靶研究中得到的中子剂量学参数进行了比较和分析。     

金属铋靶件制备研究

采用高温化学萃取法可以从反应堆辐照后的金属铋（²⁰⁹Bi）中分离²¹⁰Po。设计了两种规格的金属铋靶件,用MCNP程序模拟靶件在中国先进研究堆（CARR）中的中子注量率及核发热情况,对两种靶件的传热进行计算。用氩弧焊对金属铋靶件进行焊接,对焊接后的靶件进行密封性检查。结果表明：两种靶件在CARR反应堆中满功率60 MW运行时中子注量率最大为5.21×10¹⁴ n/cm²·s,靶件总发热量分别为1 707 W和2 220 W,经传热计算后靶件中心温度分别为163.5 ℃和191.8 ℃,低于金属铋的熔点（271.3 ℃）。焊接后的靶件经密封性检查,泄漏率小于3.2×10^-9 Pa·m³/s,金属铋靶件可用于CARR反应堆辐照制备²¹⁰Po核素。     

厚铍靶9Be(d,n)反应中子产额测量

能量在3MeV以下厚靶D-Be反应的中子产额实验数据至关重要,但较为缺乏。本工作在北京大学4.5MV静电加速器上对氘束轰击厚铍靶的中子产额进行测量。对入射氘核能量在0.35~2MeV之间的若干能量点用长中子计数管进行了0°方向中子产额、中子角分布及中子总产额的测量。与已有的测量结果和经验公式进行了比较,并拟合出氘束轰击厚铍靶中子总产额的经验公式。     

高速旋转氚钛靶系统设计和靶温度的数值模拟

给出了用于强流中子发生器的高速旋转氚钛靶系统的设计方案,并对靶的温度变化进行了数值模拟,给出了强流中子发生器的运行参数。     

用组合叠层CR-39固体径迹探测器测量加速器厚铍靶9Be(d,n)反应中子能谱

采用组合叠层CR-39固体径迹探测器实验方法测量了加速器D（d,n）反应产生的5MeV与2MeV准单能中子能谱。进而测量了入射氘离子能量为3MeV时加速器厚铍靶9Be（d,n）反应的中子能谱,与已有的飞行时间法的测量结果基本相符。在此基础上,用该法又测量了入射氘离子能量为1．5MeV时加速器厚铍靶9Be（d,n）反应的中子能谱,结果符合较低能量氘离子与厚铍靶发生9Be（d,n）的核反应的物理过程。     

Development of a finite volume inter-cell polynomial expansion method for the neutron diffusion equation

Heterogeneous nuclear reactors require numerical methods to solve the neutron diffusion equation (NDE) to obtain the neutron flux distribution inside them, by discretizing the heterogeneous geometry in a set of homogeneous regions. This discretization requires additional equations at the inner faces of two adjacent cells: neutron flux and current continuity, which imply an excess of equations. The finite volume method (FVM) is suitable to be applied to NDE, because it can be easily applied to any mesh and it is typically used in the transport equations due to the conservation of the transported quantity within the volume. However, the gradient and face-averaged values in the FVM are typically calculated as a function of the cell-averaged values of adjacent cells. So, if the materials of the adjacent cells are different, the neutron current condition could not be accomplished. Therefore, a polynomial expansion of the neutron flux is developed in each cell for assuring the accomplishment of the flux and current continuity and calculating both analytically. In this polynomial expansion, the polynomial terms for each cell were assigned previously and the constant coefficients are determined by solving the eigenvalue problem with SLEPc. A sensitivity analysis for determining the best set of polynomial terms is performed.     

快中子活化法对声空化核效应的验证

采用纯铜作为阈探测器检测声致核聚变产生的14 MeV中子。根据14 MeV中子与Cu的核反应,选择合适的放射性核素及其特征γ峰作为测量依据。中子辐照时间为50 min,经30 min和198 min冷却,NaI探测器分别测量了超声和非超声下活化铜片的511 keV特征γ峰计数,测量结果显示,采用短冷却时间可测得⁶²Cu的511 keV γ特征峰,γ峰净面积计数增量ΔC均为正值,具有统计意义,在声空化条件下核反应液体中D-T反应产生的14 MeV中子发生率大于在非声空化条件下的;采用长冷却时间可测得⁶⁴Cu的511 keV γ特征峰,ΔC均为正值,具有统计意义,在声空化条件下核反应液体中D-D反应产生的2.45 MeV中子发生率大于在非声空化条件下的。由此验证了声空化核效应（NEAC）,并初步分析了中子成核声空化核效应的机制。     

Fission Spectrum Averaged Cross Sections of Some Threshold Reactions Measured with Fast Reactor “YAYOI”

Abstract

Fission spectrum averaged cross sections of twenty one threshold reactions were measured in the core center of YAYOI which was a fast neutron source reactor. Fast neutron spectrum in the core was experimentally determined by using a set of activation foils and micro-fission counters, prior to the cross section measurement. It was found that the shape of the fast neutron spectrum was approximately the same as that of fission neutrons above about 2MeV. This fact was also supported by theoretical calculation.

Since this neutron field has scarce thermal and epithermal neutrons, measurement of nuclei produced by threshold reactions is not affected by (n, γ) reactions which are induced by thermal and epithermal neutrons. Moreover, considerably high fast neutron flux (about 5 x 10¹¹n/cm²·sec) enables to measure cross sections of small values.

The results in general agreed with the previous values obtained in a reactor core or with a fission plate within an experimental error, while they were systematically smaller by about 10% than those recommended by Fabry. The measured values are also compared with the results calculated by Pearlstein based on a statistical model.     

Neutron Energy Dependence of Excess Charge Carrier Lifetime Degradation in Silicon

Silicon p-i-n diode-neutron dosimeters were irradiated with monoenergetic neutrons having energies from 0.36 to 4.77 MeV. The variation with neutron energy of the damage coefficient for diode storage time, ?T = ?(l/Ts)/?, and diode forward voltage sensitivity at constant current, S = ?Vf/?, were determined and found to be equal. Since ?T; and S are linearly related to the damage coefficient for high level lifetime, ?T; = ?(1/?)/?, the variation with neutron energy of ?T was also determined. Theoretically, it has been hypothesized that ?T is proportional to N?r(En), the average number of active recombination centers added per cm3 per neutron of energy En incident per cm2. It was further hypothesized that N?r(En) is proportional to ?(En), the average energy given to lattice atoms per cm3 per neutron of energy En incident per cm2. The theory of Lindhard, Nielson, Scharff, and Thomsen together with published values for differential neutron-silicon scattering, was used to calculate ?(En). Comparison of theoretical and experimental results shows that ?T, S, and ?(En) do indeed have substantially the same neutron energy dependence.     

Determinations of ~(171)Er half-life and some ~(171)Tm transition energies

The neutrons have been captured by Erbium nuclei which were received by using clinical electron linear accelerator. In this experiment, the possibility of the neutron capture process has been observed because of emitted neutrons appearing in the experimental area. In particular,neutron capture of ~(170)Er nucleus has been observed. After the neutron capture of ~(170)Er nucleus, the unstable ~(171)Er has been formed and decayed into the ~(171)Tm. By using this reaction path, some transition energies of ~(171)Tm obtained from the residual activity measurements and the half-life of ~(171)Er have been determined, and they are in agreement with adopted values in the literature.     

An Experimental Study on Subcritical Assembly with Neutron Source

A new definition of Subcritical constants is introduced in this paper on the basis of the neutron balance equation. These Subcritical constants are obtained by integrating the intensity of the neutron flux over the whole volume of the Subcritical assembly with neutron source. Some other constants commonly used in conventional reactor physics can be predicted by analyzing the above mentioned Subcritical constants.

As examples of applications of this method, the infinite medium multiplication factor for natural uranium-light water system and the critical buckling for 20% enriched uranium-light water system have been determined experimentally with the use respectively of Ra-Be and spontaneous fission neutrons as the neutron source. The experimental results obtained are in comparatively good agreement with the values from exponential experiment and with theoretical values calculated by a four-factor formula.     

Determination of Large Negative Reactivity by Integral Versions of Various Experimental Methods

The large negative reactivity is measured in Semi-Homogeneous Experimental facility (SHE). Experimental methods are Sjöstrand's pulsed neutron, source multiplication and rod drop methods beside revised King-Simmons' pulsed neutron methods. Neutron detectors are placed at various points in the core region for multi-points measurement.

Usual one-point reactor model analysis resulted in the reactivity values, strongly dependent on the detector position with the increase of subcriticality. In addition, disagreements between the used experimental methods are also pointed out.

In order to overcome these difficulties due to the spatial higher harmonics and the kinetic distortion in the neutron flux distribution, an integral version analysis is applied, in which use is made of multi-points reactor model. In the analysis, space integration of the neutron counts obtained throughout the core region is made with weights of the adjoint function of fast neutrons, calculated using the two- or three-dimensional diffusion code. The negative reactivity values determined by the integral version analysis agreed well with each other within the uncertainty of ～5% in the reactivity range down to ～50 dollars.

It is concluded that all the experimental methods are adequate for precise determination of the large negative reactivity of reactor provided that the integral version analysis is utilized or that correction is made for the change of the neutron generation time using precise calculation.     

Radiochemical Determination of Neutron Capture Cross Sections of 241Am

The thermal neutron capture cross sections and the neutron capture resonance integrals of ²⁴¹Am leading to the production of the isomer ²⁴²Am and the ground-state ^242gAm were measured radiochemically by the Cd-ratio technique with neutron flux monitors of Co/Al and Au/Al alloy. Highly-purified ²⁴¹Am targets were irradiated in an aluminum capsule by using JMTR. The neutron fluxes and their epithermal neutron fractions were determined by measuring γ-rays of ⁶⁰Co and ¹⁹⁸Au. The yields of ^242mAm and ^242gAm were decided by analyzing growth and decay curves of the α-ray activity ratios ²⁴²Cm/²⁴¹Am. The resultant thermal neutron capture cross sections are 85.7 ± 6.3 b and 768 ± 58 b for ^242mAm and ^242gAm, and the resonance integrals 114±7 b and 1,694±146 b, respectively. The differences between the present results and the evaluated values by Mughabghab are 38–59%. The isomeric ratios, g/(m+g), of 0.90±0.09 for thermal neutrons and 0.94±0.11 for epithermal neutrons are, however, almost consistent with evaluated values.     

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。     

微型中子源反应堆中子谱参数测量

以Au、Zr和Fe为活化探测器,采用裸探测器法测量中国原子能科学研究院微型中子源反应堆的中子谱参数f、α、f_F和φ_th。内辐照座的α、f和f_F分别为－0.007±0.003、20.8±0.4、5.5±0.2。该方法对φ_th的测量结果与4πβ-γ符合法的一致,相对偏差小于2%。与SLOWPOKE相比,微堆有较高的α、f_F值。与已有测量数据的比较表明,微堆中子谱在很长一个时期内是稳定的,利用微堆作为中子源的k₀法中子活化分析不需中子注量率监测器,且比较器一经照射和测量后,可用于其后较长时间内所有分析的计算标准。