快中子脉冲堆弱源爆发脉冲实验理论研究

我们在应用Hansen理论研究快中子脉冲堆弱源爆发脉冲实验时发现,理论与实验结果差距非常大。对此,我们分析了原因,然后从求解系统中子数概率分布函数出发,研究和求解了系统有限长裂变链的期望值,给出了有限长裂变链期望值的求解方法和基本形式,讨论和定量计算了Godiva-Ⅱ和CFBR-Ⅱ脉冲堆有限长裂变链的期望值。然后研究了有限长裂变链导致缓发中子先驱核增值行为,结果表明,在等待爆发脉冲期间缓发中子数可发生数倍的变化。最后,在两种不同的物理假定下,对Hansen模型给予了改进。改进后的模型可以较理想地模拟Godiva-Ⅱ和CFBR-Ⅱ两个脉冲堆的实验结果,表明结果是可信的。     

稍超瞬发临界系统缓发中子发射率研究

使用蒙特卡罗直接模拟程序计算了弱源点火条件下超瞬发临界5￠的GodivaⅠ快中子脉冲堆缓发中子发射率规律和脉冲引发时刻概率分布。计算结果表明,脉冲引发时刻概率分布的计算结果与实验结果符合较好,所获得的缓发中子发射率规律可合理解释引发时刻概率分布的实验结果。     

超瞬发临界系统有限裂变链集体效应

本文以点堆模型为基础建立了集体有限裂变链模拟程序，以宏观方法研究有限裂变链在核系统内部的发展演化规律，并得到了实验结果的验证。采用该程序计算了有限裂变链的集体链长，得到了其分布规律。结果表明，平均集体链长随源强的增加而增大，随反应性的增加而减小。     

瞬发和缓发γ射线对堆内构件释热率影响的研究

为提高核电设计中反应堆堆内构件释热率计算的准确性,本文在原来MCNP外中子源模型计算方法的基础上,计算分析瞬发裂变γ对堆内构件释热率的贡献。计算结果显示,考虑瞬发裂变γ使得堆内构件的释热率增加9%～38%,离堆芯越近的堆内构件的增加值越大。另外,分析认为缓发γ对堆内构件释热率的贡献与瞬发裂变γ相当。因而反应堆堆内构件释热率计算中除了考虑中子及中子俘获所生γ的贡献,还应该考虑瞬发裂变γ和缓发γ的贡献。     

裂变碎片瞬发中子数随质量分布的研究

基于TALYS程序计算结果，我们研发了利用放中子前后裂变碎片的质量分布计算裂变碎片瞬发中子数随碎片质量分布的方法。裂变碎片瞬发中子数随裂变碎片质量分布的形状依赖于裂变产生的激发能如何在两个碎片中分配。这个问题一直以来没有得到很好的解决。     

持续裂变链引发概率的直接模拟研究

讨论了判断持续裂变链引发的理论依据,利用基于Geant4toolkit开发的蒙特卡罗直接模拟程序,计算了一系列测试几何和超瞬发临界5c/的GodivaⅠ反应堆内持续裂变链的引发概率。结果表明,蒙特卡罗直接模拟计算的结果与Mercury程序及确定性理论结果均一致。     

脉冲中子-裂变中子铀矿测井技术的蒙特卡罗模拟

脉冲中子-裂变中子铀矿测井方法（PNFN）是采用脉冲式中子源,利用-3He管中子探测器记录瞬发裂变超热中子或缓发裂变热中子,得到地层中铀矿含量信息的测井方法。利用MCNP程序模拟了不同铀含量、不同地层孔隙度地层条件下PNFN的响应,分析了瞬发裂变超热中子和缓发裂变热中子与地层铀含量和孔隙度的关系。结果表明,地层孔隙度对利用PNFN确定地层铀含量有影响,孔隙度越大,利用裂变中子直接计算得到的地层铀含量比真实含量越小。利用瞬发裂变超热中子或热中子时间衰减谱计算得到地层宏观俘获截面,对裂变中子进行校正,可以有效提高地层铀含量计算结果的准确度。     

緩发中子有效份額的表达式

中子引起可裂变物质(铀、钚)裂变时,除了产生二次瞬发中子外,还产生少量缓发中子。缓发中子的存在对于反应堆的控制有重要的意义。它直接关系到反应堆的动态特性,以及根据反应堆动态参数(中子寿命、反应性)测定物理特性的实验结果。这类实验在反应堆物理实验中占有相当重要的地位。在实际的反应堆中,由于存在一定数量的U~(238),而且U~(238)裂变生成的缓发中子份额与U~(235)裂变时生成的不同,以及由于为反应堆的有限大小所决定的中子不漏     

瞬发裂变中子铀矿测井环境影响因素的蒙特卡罗模拟研究

瞬发裂变中子铀矿测井是一种快速、可实现直接测铀并可现场定量地层岩石中铀含量的核测井新技术。为消除瞬发裂变中子铀矿测井受地层环境因素的影响,利用MCNP5对瞬发裂变中子铀矿测井的井径、井液变化进行了深入分析,通过分析干孔、含水孔在不同孔径下的超热中子与热中子数的变化关系,使用线性拟合给出了井径修正函数,并取得了满意的实验结果。     

MC法模拟核部件自发裂变中子活化瞬发γ能谱

在分析中子活化瞬发γ产生机理及瞬发γ射线强度计算方法基础上,提出了应用MCNP程序计算模拟核部件自发裂变中子活化放出瞬发γ能谱的直接模拟与分步模拟方法,对两种方法的计算结果及特点进行了比较分析。计算了模拟核部件核材料自发衰变产生的γ能谱,并与瞬发γ能谱进行了比较分析。本文结果可为核部件认证技术研究提供参考。     

The effect of delayed neutrons on the time of establishing a stable fission chain

The stochastic process of delayed neutron multiplication is considered in limited fission chains in a reactor which is supercritical with respect to prompt neutrons. Equations are obtained for the instants of distribution of the output of delayed neutron precursors and for the mean square deviation. The effect of delayed neutrons on the time of establishment of a stable fission chain is investigated. Equations are compiled and their approximate solutions are given. It is shown that for a small reactivity and with a weak source, the average time for establishing the first stable fission chain can be reduced by the delayed neutrons by a factor of 10 or more.Translated from Atomnaya Énergiya, Vol. 18, No. 6, pp. 578–583, June, 1965     

Angle and energy distributions of fission neutrons

Using the statistical model of the nucleus the angle distribution of fission neutrons is determined taking into account the anisotropy in the angle distribution of the fission fragments. The latter quantity is approximated by a simple expression of the form 1 + kcos² α. It is assumed that the neutrons are emitted isotropically in the coordinate system in which the fragment is at rest. Using U²³⁸, a calculation is carried out to determine the values of P — the ratio of the intensity of fission neutrons emitted in the direction of the incident neutrons to that of neutrons emitted perpendicularly to the incident beam — at various energy values of the primary and secondary neutrons (in the region from 1 to 10 Mev). The mean value of P in fission in U²³⁸ by neutrons characteristic of a fission spectrum is found to be approximately 1.13. The energy spectra for fission neutrons at various primary-neutron energies are also obtained. The method of calculation can also be employed in investigations of the anisotropy of neutrons produced in fission of other nuclei.     

0.57、1.0和1.5 MeV中子诱发235U裂变的99Mo产额绝对测量

用HPGe γ能谱法绝对测量了0.57、1.0和1.5 MeV中子诱发²³⁵U裂变产物⁹⁹Mo的产额,使用双裂变室测量了样品辐照过程中的裂变率,应用MCNP ⅣB模拟了铀样品中的中子能谱,并讨论了非主中子的各种来源对产额数据的影响。得到⁹⁹Mo在0.57、1.0和1.5 MeV的产额分别为6.61%、6.62%和6.28%。本工作与美国阿贡实验室的结果有15%以上的相对偏差,主要是由引用的衰变数据不同引起。对阿贡实验室数据进行校正后,本工作与阿贡实验室数据的相对偏差处于实验不确定度范围内。     

Angle and energy distributions of fission neutrons

Using the statistical model of the nucleus the angle distribution of fission neutrons is determined taking into account the anisotropy in the angle distribution of the fission fragments. The latter quantity is approximated by a simple expression of the form 1 + kcos² .It is assumed that the neutrons are emitted isotropically in the coordinate system in which the fragment is at rest.Using U²³⁸, a calculation is carried out to determine the values of P — the ratio of the intensity of fission neutrons emitted in the direction of the incident neutrons to that of neutrons emitted perpendicularly to the incident beam — at various energy values of the primary and secondary neutrons (in the region from 1 to 10 Mev). The mean value of P in fission in U²³⁸ by neutrons characteristic of a fission spectrum is found to be approximately 1.13. The energy spectra for fission neutrons at various primary-neutron energies are also obtained. The method of calculation can also be employed in investigations of the anisotropy of neutrons produced in fission of other nuclei.In conculations the authors wish to thank V. K. Sauleva for setting up the prograsm for the electronic computer.     

从广义自持链式反应观点看加速器驱动系统

用广义自持链式反应的观点探讨了加速器驱动系统 (ADS)的基本内涵。认为次临界反应堆、质子加速器和靶所组合的整体仍可看成一个 (临界的 )自持链式反应堆。这个反应堆不同于通常临界反应堆的特点是每次裂变后的二次中子不仅包含裂变释放的中子而且还包含部分裂变释能 (通过质子加速器及靶 )所转换的中子。正是有了这些附加中子 ,使得加速器驱动系统每次裂变的有效二次中子数增加了。一个ADS系统能够稳定运行的条件是ADS的次临界堆和加速器能够相互匹配使得ADS系统的有效二次中子数达到这样的水平 ,以致在ADS系统内能够形成自持的中子链式反应。因此尽管ADS的反应堆部分是次临界的 ,但从ADS整体来看只要质子加速器与次临界反应堆匹配得当 ,ADS系统是可以像通常临界反应堆那样 ,维持自持的链式反应的 (或临界的 )。给出了ADS系统维持自持链式反应的匹配条件 (广义临界条件 )。最后根据ADS系统的特点探讨了ADS在核废物处理 (嬗变 )、提高核燃料增殖效率及核能开发中的作用。     

Improvements in the description of the multiplicities of the prompt neutrons emitted in fission

Various parameterizations of the multiplicity distribution of prompt neutrons from nuclear fission are investigated. The currently accepted formalism is found to fail for those fission processes which release on average a low number of neutrons. Based on realistic assumptions regarding binary fission, an improved multiplicity distribution formula gives much better agreement with the experimental data. Explicit details of the practical usage of the new formalism are presented. Finally, it is shown that for spontaneously fissioning isotopes, it is possible to calculate the variance of the neutron multiplicity distribution from a knowledge of the mass yield and the variances of the neutron multiplicity distributions of the individual fission fragments.     

随机中子输运方程数值解与近似解的对比分析

在描述具有弱源的有限增殖系统引发持续裂变链概率时,随机中子输运理论点堆模型近似解和考虑中子位置和速度方向(r,v)相空间模型的数值解的形式不同,用一个与球形堆半径r相关的改进点堆模型近似解与Sn方法数值解进行比较,由此可以看出两者的差异及非线性项的影响程度.     

Investigation of the statistical distribution of spontaneous U238 fission acts according to the energies of the two fragments

Using an apparatus consisting of a double pulse ionization chamber, two amplifying channels, a coincidence circuit and a two-beam pulsed oscillograph, we measured the kinetic energy of each of the two fragments formed during nuclear fission. 780 cases of spontaneous U²³⁸ fission and about 4500 cases of U²³⁵ fission by slow neutrons were recorded. Using these data we plotted the statistical distributions of acts of spontaneous fission and fission with slow neutrons according to the energies of the two fragments. These distributions gave the following curves for the two fission methods: a) the energy spectrum of all fragments and that of light and heavy fragments separately; b) the distribution of acts according to the total kinetic energy of the two nuclear fragments; c) the distribution of fission fragments according to mass; d) the relation of the average total kinetic energy of the two fragments to the ratio of their masses; and e) the relation of the average kinetic energy of light and heavy fragments to their total kinetic energy. The results of the two methods of fission are quite similar. It was found that the total kinetic energy of the two fragments in spontaneous U²³⁸ fission was, on the average, 4 Mev less than the energy in U²³⁵ fission by slow neutrons.     

Experimental study of neutron counting in a zero-power reactor driven by a neutron source inherent in highly enriched uranium fuels

Even a zero-power reactor core containing highly enriched uranium has a weak neutron source inherent in uranium 235, and consequently, a neutron counter placed closely to the core without external neutron source registers a certain counting rate. The study of the counting is very important for zero-power reactor physics experiments with a high precision. In this experimental study, first, at a shutdown state of the UTR-Kinki reactor without start-up neutron source, a pulse height distribution of output signals from a neutron proportional counter was measured to confirm that these signals resulted from neutron detections. At several subcritical states of the UTR, then, the Feynman-α analysis was carried out to confirm that the neutrons detected by the counter must be fission neutrons multiplied by fission chain reactions. The correlation amplitude measured in the Feynman-α analysis was much higher than that measured in a previous drive by start-up source. Further, it was also confirmed that the subcriticality dependence of neutron counting rate followed the source multiplication formula. This feature indicated that the one-point model was very successful in the subcritical range including the shutdown state.     

Neutron Emission,Mass Distribution and Energetics in 14.5 MeV-Neutron Induced Fission of Uranium-238

A study is performed on 14.5 MeV-neutron induced fission of ²³⁸U by means of three-parameter experiment in which the energies of both fragments and the time-of-flight of one fragment are measured. A mosaic-arrayed surface barrier detector of large sensitive area is used at the remote end of a flight tube. The pre- and post-neutron-emission fragment mass distributions are obtained, together with the average total kinetic energy of fragment as a function of its mass. The average number of neutrons emitted from an individual fragment and the average total number of emitted neutrons are also derived as a function of fragment mass. The results agree well with those calculated by the method developed in our laboratory for medium-excitation fission. The average number of emitted neutrons and the mass distribution of fission fragment are derived for the respective reactions of first-, second- and third-chance fission.