Numerical aspects in the study of neutron propagation

The study of the propagation of localized source pulses is an important problem in nuclear reactor physics and in the analysis of pulsed experiments. In a previous work the characterization of the transport operator has been performed and the physical distortions introduced by approximate models, such as spherical harmonics, have been evidenced by means of an analytical asymptotic approach. In this paper, the S_N and P_N−1 time-dependent models are numerically discretized in space and solved analytically in time, in order to evidence the error associated with discretization schemes.     

Optimization study on neutron spectrum unfolding based on the least-squares method

The response functions and pulse height spectrum(PHS) of a 2'×2' BC501A detector were obtained through a general-purpose Monte Carlo simulation toolkit,Geant4.A relatively simple but effective method was adopted to unfold the PHS.Recommendations regarding the response matrix were proposed to optimize the unfolding results.The results indicate that the accuracy of the unfolding can be greatly improved using many incident neutrons with a wide energy range,a proper energy interval,and an appropriate channel width of the response matrix.The above-mentioned method was verified by unfolding three different types of simulated spectrum,the results of which are in good accord with the simulated distribution.     

中子辐照GaAs的变温正电子寿命研究

用正电子湮没寿命谱研究中子辐照前后半绝缘ＧａＡｓ在等时退火过程中的缺陷行为．发现辐照后的样品大约从７０℃起空位－填隙子对和空位－反格点缺陷开始复合．５００℃前空位和空位团可消除．７４０℃后可能会重新产生缺陷。ＶＡｓ是一种浅势阱，对正电子的束缚能大约为０．０３１－０．０３２ｅＶ。     

Study of the neutron multiplication effect in an active neutron method

The neutron multiplication effect appears when an item contains large amounts of nuclear material. The neutron multiplication effect in this paper means the effect of subsequent fission reactions which are caused by fission neutrons produced by interrogation neutrons from a neutron generator. The previous active neutron method could not distinguish between first-fission and subsequent-fission neutrons and might overestimate the amount of nuclear material. However, the neutron multiplication effect in the active neutron method has not been adequately investigated. We discuss the evaluation method of the multiplication effect in the fast neutron direct interrogation method, one of the active neutron methods, using simulations with the Monte Carlo code MVP and experiments involving uranium waste drums. The first-generation neutrons from an external neutron source generate fission neutrons called second-generation neutrons, the second-generation neutrons generate third-generation neutrons, and so on. This study supposes that the neutron multiplication effect is mainly caused by the third-generation neutrons under the condition that the fourth-generation neutrons are much fewer. This paper proposes a correction method for the neutron multiplication effect in the measured data.     

反应堆压力容器快中子注量计算方法研究

依据美国核管会有关压力容器中子注量确定方法的管理导则,在二维SN程序DORT的基础上,通过自行开发X-Y与R-θ坐标系间的中子源转换程序,建立起了一套基于综合法的压力容器三维快中子注量计算方法。NUREG/CR-6453和NUREG/CR-6115基准问题检验结果表明,本文所建立的方法能够提供满足工程精度要求的压力容器注量空间和能谱分布。     

The asymptotic behavior of neutron waves in small systems and its application to the investigation of solutions of the neutron transport equation

The neutron wave propagation method has been applied in small graphite blocks to study the axial propagation of the neutron disturbance and the transverse wave propagation. No evidence of non-asymptotic behavior within the limits of the wave frequencies investigated was found, and an effect of transverse wave propagation which points to a frequency-dependent complex transverse buckling has been demonstrated for the first time.     

The study of neutron diffusion in beryllium by a pulse method

A pulsed-neutron-source method was used to determine to a high accuracy the diffusion and nuclear characteristics of beryllium: the diffusion coefficient, the diffusion and transport lengths, the mean lifetime, the effective absorption cross section, and the diffusion-cooling coefficient.The article discusses the reasons for the discrepancies between the values obtained by different authors for the diffusion-cooling coefficient.Translated from Atomnaya Énergiya, Vol. 16, No. 3, pp. 224–228, March, 1964     

Space and energy dependent theoretical formula for the third order neutron correlation technique

We have studied measurement techniques of subcriticality by using the neutron correlation techniques. Among various techniques, we have paid attention to the third order neutron correlation technique, which was proposed by Furuhashi originally and utilizes the third order fluctuations of neutron counts. By using this technique, we can obtain the absolute value of subcriticality with the aid of factorial moments of fission neutron ν and external source neutron q, without prior knowledge of the prompt neutron decay constant at a reference state. However, in order to apply the third order neutron correlation technique to actual experiments, we must consider the spatial and neutron energy effects in this technique. For this purpose, we derived more generalized theoretical formulas of the third order neutron correlation technique that took account of the spatial and neutron energy effects by virtue of the α-eigenfunction expansion. Making use of the newly derived formulas, we propose the practical formulas and the experimental procedures based on the fundamental mode approximation.     

中子测井与天然气探测技术

简单介绍了中子测量与地层含氢指数及地层孔隙度的关系,介绍了地层含气对中子测量的影响以及泥浆侵入对中子在含气地层响应特征的影响.分析了中子测井仪器长短源距探测器受泥浆侵入影响的差异,给出了通过对中子仪器长短源距计数率校正消除泥浆侵入影响的方法.通过实际资料验证,取得了比较好的结果.     

A new three-dimensional method of characteristics for the neutron transport calculation

The method of characteristics (MOC) is a very flexible and effective method for the neutron transport calculation in a complex geometry. It has been well developed in two-dimensional geometries but meets serious difficulty in three-dimensional geometries because of the requirements of large computer memory and long computational time. Due to the demand related to the advanced reactor design for complex geometries, an efficient and flexible three-dimensional MOC is needed. This paper presents a modular ray tracing technique to reduce the amount of the ray tracing data and consequently reduce the memory. In this method, the object geometry is dissected into many cuboid cells by a background mesh. Typical geometric cells are picked out and ray traced, and only the ray tracing data in these typical cells is stored. Furthermore, the Coarse Mesh Finite Difference (CMFD) acceleration method is employed to save computing time. Numerical results demonstrate that the modular ray tracing technique can significantly reduce the amount of ray tracing data, and the CMFD acceleration is effective in shorting the computing time.     

Applicability of the enhanced neutron current method for non-convex fuel shapes

The enhanced neutron current method is applied for the evaluation of the background cross-section for fuels with non-convex shape. The evaluated background cross-sections are compared with those obtained by the conventional method (the equivalence theory) commonly used in the resonance calculation. However, the Dancoff factor for a fuel with non-convex shape is not directly obtained by the traditional method currently used in the resonance calculation. Therefore, a new method for the evaluation of the Dancoff factor is also proposed, which utilizes the variation of moderator density and is applicable to the non-convex shapes. The test calculations for fuels with non-convex shape indicate that the background cross-section obtained by the enhanced neutron current method and the conventional method agree well with each other. Since the enhanced neutron current method can be applicable not only for the convex shapes but also for non-convex shapes, it will be useful for the resonance calculations in complicated geometry.     

A new method for the evaluation of neutron elastic transformation matrices

A new method for the evaluation of the center-of-mass to laboratory co-ordinates system transformation matrix is developed, by which each element of the matrix is represented as a finite sum of terms depending on the mass number of the scattering nucleus. At present, the elastic transformation matrices given in the ENDF/B files are evaluated by the Amster method, by which each element of the transformation matrix is represented as an infinite sum of terms. A comparison is made between the values of the elements of the elastic transformation matrix of ⁶Li obtained by the new method and those given in the ENDF/B-V file, and some discrepancies are revealed.     

Development of the nodal collocation method for solving the neutron diffusion equation

The nodal collocation method is a new technique for the discretization of the multidimensional neutron diffusion equation where the solution sought is expressed in the form of tensorial expansions of Legendre polynomials defined over homogeneous parallelepipeds. In this study, we have truncated the tensorial expansions using the serendipity approximation in an attempt to reduce the total number of unknowns and improve the effectiveness of the discretization. The remaining Legendre coefficients are then determined in order to preserve selected moments of the neutron conservation equation in each parallelepiped. This approach allows a variable order of convergence without sacrificing the consistency peculiar to full tensorial expansions and generates matrix systems which may be resolved by an Alternating Direction Implicit algorithm. Furthermore, we have proved that the linear nodal collocation method and the mesh centered finite difference method are equivalent. Validation results are given for the IAEA 2-D and 3-D benchmarks and for a 2-D representation of a pressurized water reactor (PWR).     

Generalized analysis method for neutron resonance transmission analysis

Neutron resonance densitometry (NRD) is a non-destructive analysis method, which can be applied to quantify special nuclear materials (SNM) in small particle-like debris of melted fuel that are formed in severe accidents of nuclear reactors such as the Fukushima Daiichi nuclear power plants. NRD uses neutron resonance transmission analysis (NRTA) to quantify SNM and neutron resonance capture analysis (NRCA) to identify matrix materials and impurities. To apply NRD for the characterization of arbitrary-shaped thick materials, a generalized method for the analysis of NRTA data has been developed. The method has been applied on data resulting from transmission through thick samples with an irregular shape and an areal density of SNM up to 0.253 at/b (≈100 g/cm²). The investigation shows that NRD can be used to quantify SNM with a high accuracy not only in inhomogeneous samples made of particle-like debris but also in samples made of large rocks with an irregular shape by applying the generalized analysis method for NRTA.     

降低闪烁探测系统中子灵敏度的方法初探

针对脉冲辐射场高强度裂变中子测量的特点,设计了降低灵敏度的狭缝闪烁探测系统,初步研究了系统的2.5MeV中子灵敏度。实验结果表明：由狭缝屏蔽准直器和闪烁探测器组成的新型狭缝闪烁探测系统,可使传统的直照探测模式的下限灵敏度降低一个量级,拓宽了闪烁探测器在脉冲辐射场中子测量中的应用。     

Effect of235U fission on the propagation velocity of the neutron fission wave

Institute of High-Energy Physics, Academy of Sciences of the Kazakh SSR (IFVÉ AN KazSSR). Translated from Atomnaya Énergiya, Vol. 71, No. 4, pp. 309–311, October, 1991.