Pulsed Neutron Intensity from Rectangular Shaped Light Water Moderator with Fast-Neutron Reflector

With a view to enhancing the thermal-neutron intensity obtained from a pulsed neutron source, an experimental study has been made to determine the optimum size of a rectangular shaped light water moderator provided with fast neutron reflector of beryllium oxide or graphite, and decoupled thermal-neutronically by means of Cd sheet. The optimum dimensions for the moderator are derived for the neutron emission surface and the thickness, for the cases in which the neutron-producing target is placed beneath the moderator (“wing geometry”) or immediately behind the moderator (“slab geometry”). The major conclusions drawn from the experimental results are as follows. The presence of the Cd decoupler inserted between the moderator and reflector prevent the enhancement of thermal-neutron emission time gained by the provision of reflector. with a graphite reflector about 14 cm thick, (a) the optimum area of emission surface would be 25x25 cm² for wing geometry and still larger for slab geometry, and (b) the optimum moderator thickness would be 5.5 cm for slab geometry and 8.5 cm for wing geometry. It is thus concluded that a higher neutron emission intensity can be obtained with slab than with wing geometry provided that a large emission surface can be adopted for the moderator.     

Neutronic Optimization of Premoderator and Reflector for Decoupled Hydrogen Moderator in 1 MW Spallation Neutron Source

For a decoupled liquid-hydrogen moderator, optimization studies have been performed on a premoderator and reflector materials (Pb, Be, Fe and Hg) together with several decoupling energies to realize a higher neutronic performance. The result indicated that, among four reflector materials mentioned above, the best neutronic performance could be obtained by adopting a Pb reflector with an optimized premoderator and an appropriate decoupling energy.     

Neutronics of Polyethylene Thermal Moderator of Wing and Slab Geometries on Pulsed Neutron Source

Pulse width and intensity of thermal neutrons emitted from a polyethylene moderator on a pulsed neutron source are compared in two geometries of a moderator and target, namely, slab (radial type) and wing geometries (tangential type).

The pulse width for a 10cm thick moderator of the wing geometry, at which thickness saturated beam intensity is achieved, corresponds nearly to that of a 4 cm thick moderator of the slab geometry, although the maximum beam intensity for the wing geometry is some-what less than that of slab geometry. Analysis of figures of merit indicates that the wing geometry can be adopted in scattering experiments which require neutron beams that are not contaminated much by other radiations, even if it causes slight deterioration of neutronic characteristics in comparison with the slab geometry.     

Optimization of Grooved Thermal Moderator for Pulsed Neutron Source,and Its Characteristics

A grooved moderator having grooves on the neutron emanating surface of the moderator has been proposed for the purpose of increasing the neutron intensity of the pulsed neutron source. Experimental studies on polyethylene grooved moderators are performed, and the effects of the groove, the optimum dimensions and the influence of the reflector on the optimum condition are examined with a view to enhancing the emanating thermal neutron intensity. The pulse shapes of emanating neutrons are also measured and the figures of merit of the grooved moderator are discussed.

The experimental results reveal that the optimum condition is determined by the balance of increase of intensity from the grooves and the decrease from the fins. The optimum dimensions for the moderator with a graphite reflector are as follows. The fin width is less than 1 cm, the fin length 8~10 cm and the fin area about 60% of the emanating surface. These values does not depend strongly on whether the moderator is covered by the reflector or not. This size of grooved moderator increases the emanating neutron intensity by about twice at the neutron energy of 25 meV, as compared with the rectangular shaped moderator of 5 cm thick, although it broadens the pulse width considerably. Analysis of the figures of merit suggests that although the grooved moderator is not useful for the usual neutron scattering experiments, it may be of value in experiments having a resolution which is not. very sensitive to the pulse width.     

热中子俘获瞬发γ射线强度计算

简要介绍了热中子俘获发γ射线强度的计算方法,并以实例进行说明,还给出了有关物理自洽检验的方法及其讨论。     

Pulsed Cold Neutron Source of Solid Methylbenzene

Lithium carbonate pellets are frequently used for estimation of tritium production rate in irradiated samples in fusion blanket neutronics experiment and the activity is measured by liquid scintillation counting technique. In this measurement, it is essential to solve the lithium carbonate pellet as much as possible and to mix the pellet solution into scintillation cocktail homogeneously at stable condition. For this purpose, a novel binary-acid method has been developed to solve lithium carbonate and to mix the pellet solution into scintillation cocktail. High solubility is attained by adopting two acids, HNO₃ and CH₃COOH, and a good compatibility of the pellet solution with scintillator is obtained by emulsion cocktail resulting in high counting efficiency. Defining a product of dissolved mass and counting efficiency as a Figure of Merit (FOM), the present method has higher FOM value than the conventional method and is extremely simple in a sample preparation procedure. In the present work, solubility, compatibility and counting efficiency were systematically examined for different mixing ratios of two acids, and the condition for a maximum FOM was determined. The FOM value of a factor of two higher than the conventional method was finally attained.     

Measurements and Calculations of Angular Flux Spectra Emitted from Lithium and Graphite Slabs with D-T Neutron Source

Angular dependent flux spectra from slab assemblies (lithium and graphite) were measured to test nuclear data and calculational methods for D-T fusion reactor neutronics. The collimated 14 MeV neutron source could be applied by the use of an associated particle method and the neutron spectra from 14 to 2 MeV were observed with TOF technique. The measured spectral pattern was dependent on the anisotropy of secondary neutrons emitted from both the elastic and the non-elastic scattering for 14 MeV neutrons. As for the numerical calculations, one-dimensional discrete ordinates transport codes (ANISN and NITRAN) were used. The multigroup cross sections processed with SPTG4Z from ENDF/B-IV were used as common nuclear data base. The problems of calculational methods and nuclear data were discussed in comparison with the experimental data and it was clarified that sufficient nuclear data of angular dependent cross sections for the non-elastic scattering have not been available in ENDF/B-IV and that the anisotropy of the scattering could not be calculated with ANISN which utilized the scattering kernel generated by incorrect treatment of scattering kinematics in the processing code. However, good agreement between the measurements and calculations was obtained by the use of NITRAN system with the appropriate processing codes of inelastic scattering anisotropies. It was shown that the NITRAN system was useful for anisotropic neutron transport calculations.     

252Cf随机脉冲源方法测量球形钚系统瞬发中子衰减常数

²⁵²Cf随机脉冲源方法由早期的重复脉冲源方法演变而来,是测量核系统瞬发中子衰减常数α的有效方法。采用该方法测量了钚球装配31 mm、29 mm厚钢反射层核系统的α,在有效信号和噪声的比例为1∶1的情况下,得到的瞬发中子衰减谱信噪比为7∶1,最小二乘拟合结果依次为2.25 μs^-1和3.00 μs^-1,拟合误差为±0.02 μs^-1。与Rossiα方法的测量结果进行了比较,两种测量方法的结果差异小于1.3%。     

Effects of Spatial Harmonics on Thermal Neutron Spectra Measured by Time-of-Flight Method

Spatial effect—the variations in space of transient neutron flux noted in pulsed neutron experiments, and which is ascribable to the occurrence in moderator of spatial harmonics—is studied on the measured flight time spectra of neutrons reentering the atmosphere from parametrically varied depths of a graphite moderator. Measurements were made using moderator slabs of two different axial lengths, to examine the influence of differences in moderator size on the spatial effect. The master equation representing the scalar neutron flux as function of time and energy is solved by a semi-analytic method that takes account of spatial harmonics. Simulation of the solution thus obtained of the master equation proved to reproduce the measured flight time spectra with good accuracy. It was indicated that the spatial effect on the thermal neutron spectrum is stronger with a larger than with a smaller moderator slab. The neutron temperature in graphite moderator was calculated by formula fitted in Maxwellian distribution to the peaks of the thermal neutron spectra. It was indicated that in a graphite moderator of 120 cm axial length, spatial harmonics caused a variation of more than 40 K in neutron temperature between depths in moderator differing by 30 cm.     

Neutronics of Rectangular Parallelepiped Polyethylene Moderator in Wing Geometry for Accelerator Based Thermal Neutron Source

Numerical and experimental studies of the wing geometry moderator are performed in order to examine (a) the effects of the target position and the moderator thickness on the beam intensity and on the pulse shapes emitted from a polyethylene thermal moderator, and (b) the optimum thickness of the moderator.

The beam intensity emitted from the moderator is expressed by an integration of the product of the source neutron distribution and the beam intensity produced by a unit intensity point source in the moderator. By applying this expression mechanism is analyzed for the optimum target position and the saturation phenomena of the intensity and the pulse width emitted from the moderator. The optimum target position is at about 2 cm from the neutron emission surface for moderators thicker than 4 cm and at about half of the moderator thickness for thinner ones. The intensity and the pulse shapes emitted from the moderator vary little if the target distance is varied around the optimum one and become close to the saturated ones at about 8 cm thickness. It is indicated by the analysis of figures of merit that a moderator of 4~6 cm thickness is optimum.     

基于可移动紧凑加速器D-T中子源的热中子层析实验研究

中子照相是十分重要的无损检测方法之一,尤其是针对含氢材料、同位素等的无损检测,中子照相技术具有其他射线成像不可比拟的优势。中国工程物理研究院核物理与化学研究所基于紧凑型D-T中子源,研发了可移动中子成像检测仪,成功实现了热中子照相和快中子照相实验检测。为确定基于该装置开展热中子层析检测的可行性,本文进行了数值模拟计算,利用该仪器开展了针对轻重材料模拟件的热中子层析成像实验,利用采集的181幅投影图像,在图像信噪较低和采集幅数较少条件下,成功重建了铝和聚乙烯材料包裹下的0.2 mm直径的钆丝。     

研究堆冷中子源冷包的强度分析

利用ANSYS软件包对某研究堆冷中子源（CNS）的冷包在内外压共同作用下的强度进行分析。分析结果表明，冷包的原设计不甚合理，冷包的局部应力超过了设计应力。变更原结构的几何尺寸后进行了进一步计算分析，并提出了相应的补强方案。改进后冷包的计算结果表明，补强后的结构很好地满足设计要求。研究结果为冷包的实际工程设计提供了依据。     

Measurements of Neutron Capture Cross Section of 237Np for Fast Neutrons

The neutron capture cross section of ²³⁷Np has been measured for fast neutrons supplied at the center of the core in the Yayoi reactor. The activation method was used for the measurement, in which the amount of the product ²³⁸Np was determined by γ-ray spectroscopy using a Ge detector. The neutron flux at the center of the core calculated by the Monte Carlo simulation code MCNP was renormalized by using the activity of a gold activation foil irradiated simultaneously. The new convention is proposed in this paper to make possible a definite comparison of the integral measurement by the activation method using fast reactor neutrons with differential measurements using accelerator-based neutrons. “Representative neutron energy” is defined in the convention at which the cross section deduced by the activation measurement has a high sensitivity. The capture cross section of ²³⁷Np corresponding to the representative neutron energy was deduced as 0:80 ± 0:04b at 214 ± 9 keV from the measured reaction rate and the energy dependence of the cross section in the nuclear data library ENDF/B-VII.0. The deduced cross section of ²³⁷Np at the representative neutron energy agrees with the evaluated data of ENDF/B-VII.0, but is 15% higher than that of JENDL-3.3 and 13% higher than that of JENDL/AC-2008.     

Influences of Neutron Source Spectrum and Thermal Neutron Scattering Law Data on the MCNPX Simulation of a Cyclotron-Based Neutron Field for Boron Neutron Capture Therapy

In our previous study, the simulation of a cyclotron-based neutron field for boron neutron capture therapy (BNCT) using a (p,n) spallation source with the MCNPX code was validated through measurements of the neutron energy spectrum behind the moderator assembly and the thermal neutron distribution in an acrylic phantom using reaction rates of ¹⁹⁸Au. These validations showed that the simulation generally well reproduced the measurements. However, some discrepancies between the measurements and the calculation remained for clinical trials. In this paper, we investigated the influences of neutron source spectrum and thermal neutron scattering law data in the simulation to resolve those discrepancies. We also compared measured and calculated neutron doses behind the moderator assembly with results obtained using a tissue equivalent proportional counter. We clarified that the neutron source spectrum calculated using the LA150 data led to the overestimation of high-energy neutrons in a phantom, but this overestimation did not significantly affect the neutron dose distribution in a phantom, because a dominant part of the absorbed dose is due to neutrons of energies below 1MeV. The study of the influence of neutron scattering law data in a phantom also indicated that the use of selected S(α,β) data led to an improvement in the simulation of thermal neutron behavior.     

Measurement of Effective Capture Cross Section of Americium-243 for Thermal Neutrons

The effective capture cross section of ²⁴³Am for thermal neutrons was measured with an activation method. A sample of ²⁴³Am was irradiated for 10 hrs at Kyoto University Reactor, KUR. After the irradiation, the sample was cooled for one month. In the cooling time, ^244mAm and ^244gAm produced by the irradiation decayed out to ²⁴⁴Cm. The α rays emitted from ²⁴³Am and ²⁴⁴Cm were measured with a silicon surface barrier detector. The neutron flux at the irradiation position was monitored using Au/Al and Co/Al wires. The effective capture cross section was deduced as 174.5±5.3b from the α-ray counts and the neutron flux. The quantity r√T/T₀ in Westcott's convention was 0.037±0.004. The present result was compared with the effective capture cross sections from the JENDL-3.3 and the Mughabghab evaluations.     

The Effect of Cylindrical Holes in a Moderator on the Passage of Virgin Neutrons from a Point Source

The mean free path and the mean square free path of 14 MeV neutrons, generated isotropically from a d-t point source located at the center of the moderator region of a square lattice composed of parallel circular cylindrical holes in water, are calculated for various void ratios. Very interesting inequalities between the axial and transversal components of the mean free path and the mean square free path are found, which contradict with the results of Behrens for homogeneous neutron sources. We may conclude that the effects on the free path lengths of the position of the point source located in an anisotropic and heterogeneous medium cannot be neglected.     

三维方孔硅探测器的热中子探测影响研究

采用基于Monte-Carlo方法的Geant4程序对填充~6LiF中子转换材料的三维方孔硅探测器的热中子探测进行了模拟,研究了探测器结构与探测效率、能量沉积谱的关系。探究了方孔截面尺寸、孔间距、孔深度、系统最小可探测限(LLD)等参数对热中子探测效率的影响。研究结果表明,探测效率随截面尺寸或孔间距的增大先增加后减小,随孔深度的增加而增大,直到一个极限值。经优化结构参数,在LLD为300 keV的情况下,孔间距大于6μm的三维方孔硅探测器的探测效率受LLD的影响较小。理论上,三维方孔硅探测器的最佳尺寸为孔间距6μm、孔截面尺寸30μm、孔深度1 mm,其探测效率可达59.5%。     

流体物性对冷中子源冷包截面含气率的影响模化研究

采用差压法测出了冷中子源模拟冷包中环形通道内不同层面的平均含气率,发现液相的表面张力与密度的比值是决定截面含气率的主要物性因素;此比值越大,含气率越低。氟利昂113的表面张力与密度的比值比液氢的相应值低。故可以选用氟利昂113作为模拟工质对中国先进研究堆冷中子源中气液两相氢循环的含气率进行模拟研究,且试验结果将偏于保守。     

Temperature and Concentration Dependence of Tritium Release Behavior from Solidified LiF-PbF2 Pellet Irradiated by Thermal Neutron

Thin pellets of the LiF-PbF₂ system with no porosity were irradiated by thermal neutron. Then release experiment of tritium produced in the pellets was carried out at some constant temperatures. Thus diffusion coefficients of tritium in the LiF-PbF₂ system were estimated to obtain an activation energy of diffusion, which was found to be 0.51 eV at LiF mole fraction of 0.467. It was also observed that decrease of LiF concentration in the pellet brings increase of diffusion coefficient of tritium.