A hybrid method for the simulation of radionuclide contaminant plumes in heterogeneous unsaturated formations

The numerical simulation of multiphase flow and contaminant transport problems that take place around nuclear waste repositories is an important tool for selecting appropriate sites for these facilities. The goal of this paper is to present an accurate and efficient simulation tool to approximate flows in heterogenous unsaturated hydrogeological environments. We employ a hybrid method that combines a Eulerian finite volume approximation scheme for the underlying water–air flow problem with a locally conservative Lagrangian particle tracking method to approximate the transport of a radionuclide within the water phase. The Lagrangian scheme is computationally efficient and theoretically free of numerical diffusion. Numerical simulations showing the effectiveness of the new procedure are presented and discussed.     

A probabilistic study of the effect of retardation factor uncertainty using a compartment model for radionuclide release into the biosphere

The effect of retardation factor uncertainty with reference to the results of a compartment model, representing radionuclide release into the biosphere from a disposal facility, is studied through the application of polynomial chaos theory. We review the derivation of the working equations and apply the polynomial chaos expansion to these equations. The uncertainty in the retardation factor typically covers several orders of magnitude and stems from uncertainties in the distribution coefficient, these large uncertainties make any quantitative analysis of radionuclide release highly problematic. In this paper, we assume that the uncertainty in the retardation factor is fully described by a log-uniform distribution. We compare results using polynomial chaos against a semi-analytical solution for a short decay chain and present numerical results from polynomial chaos applied to a longer generic decay chain.     

CFD技术与SACTI模型在自然通风冷却塔雾羽局地环境影响模拟中的应用比较研究

本文利用CFD技术和SACTI模型对山东某电厂冷却塔附近流场以及冷却塔雾羽抬升以及漂滴与盐的沉积对局地环境影响进行了模拟计算,并与现场试验结果进行对比分析验证。结果表明,二者模拟结果与现场试验数据基本吻合, SACTI模型预测漂滴与盐沉积的最大距离出现在冷却塔附近200 m范围内,CFD技术模拟漂滴与盐沉积的最大距离出现在下风向500 m左右。SACTI模型预测漂滴与盐地面最大沉积量约为CFD技术模拟地面最大沉积量的3倍,主要原因是由于CFD在模拟过程中考虑的冷却塔局地环流以及周围建筑物对漂滴沉积的影响。     

A stochastic model for neutrons simulation considering the spectrum and nuclear properties with continuous dependence of energy

In this work we developed a stochastic model to simulate neutron transport in a heterogeneous medium, considering continuous neutron spectra and the nuclear properties with its continuous dependence on energy. This model was implemented using the Monte Carlo method for the propagation of neutrons in different media. Due to restrictions with respect to the number of neutrons that can be simulated in reasonable computational time we introduced a variable control volume together with (pseudo-) periodic boundary conditions in order to overcome this problem. This study allowed a detailed analysis of the influence of energy on the neutron population and its impact on the life cycle of neutrons. From the results, even for a simple geometrical arrangement, we can conclude that there is need to consider the energy dependence and hence defined a spectral effective multiplication factor per Monte Carlo step.     

A CFD simulation process for fast reactor fuel assemblies

A CFD modeling and simulation process for large-scale problems using an arbitrary fast reactor fuel assembly design was evaluated. Three-dimensional flow distributions of sodium for several fast reactor fuel assembly pin spacing configurations were simulated on high performance computers using commercial CFD software. This research focused on 19-pin fuel assembly “benchmark” geometry, similar in design to the Advanced Burner Test Reactor, where each pin is separated by helical wire-wrap spacers. Several two-equation turbulence models including the k-? and SST (Menter) k-ω were evaluated. Considerable effort was taken to resolve the momentum boundary layer, so as to eliminate the need for wall functions and reduce computational uncertainty. High performance computers were required to generate the hybrid meshes needed to predict secondary flows created by the wire-wrap spacers; computational meshes ranging from 65 to 85 million elements were common. A general validation methodology was followed, including mesh refinement and comparison of numerical results with empirical correlations. Predictions for velocity, temperature, and pressure distribution are shown. The uncertainty of numerical models, importance of high fidelity experimental data, and the challenges associated with simulating and validating large production-type problems are presented.     

A proposal on accuracy estimation method for the sampling-based uncertainty analysis with Monte Carlo simulation technique

The sampling-based uncertainty analysis method is a stochastic approach to estimate response uncertainties caused by the uncertainty in the input parameters. Conventionally, to minimize the effects caused by the Monte Carlo stochastic uncertainty, lots of particle histories have been used for the uncertainty analysis. However, this can cause inefficiencies in the uncertainty analysis. To optimize the calculation efficiency, how the Monte Carlo stochastic uncertainty influences the response uncertainty should be clearly verified. In this study, a method to estimate the accuracy of the response uncertainty is proposed by introducing a standard error and an error propagation theory. Using the proposed method, response uncertainties and standard errors of the multiplication factors for three benchmark problems are evaluated by the Monte Carlo method. Our results show that the proposed method can accurately estimate the accuracy of the response uncertainty caused by the input uncertainty in using the Monte Carlo simulation method. The proposed method can be directly utilized to estimate the accuracy of the sampling-based uncertainty analysis using the Monte Carlo simulation method. Also, it is expected that the proposed method will contribute to an increase in the calculation efficiency in the sampling-based sensitivity and uncertainty analysis.     

Development of a hybrid method to improve the sensitivity and uncertainty analysis for homogenized few-group cross sections

In the framework of two-step method of reactor core calculation, few-group homogenized cross sections generated by lattice-physics calculations are key input parameters for the three-dimensional full-core calculation. Conventional method for few-group cross-sections sensitivity and uncertainty (S&U) analysis related to the nuclear data was performed based on the effective self-shielding cross sections instead of the continuous-energy cross sections, which means resonance self-shielding effect (implicit effect) is neglected. Furthermore, the multi-group covariance data is generated from the continuous-energy cross sections. Therefore, in order to perform S&U analysis with respect to the continuous-energy cross sections for both accuracy and consistency, a hybrid method is proposed in this paper. The subgroup-parameter sensitivity-coefficients are calculated based on the direct perturbation (DP) method. The sensitivity-coefficients of the effective self-shielding cross sections and the responses (k_eff and few-group homogenized cross sections) are calculated based on the generalized perturbation theory (GPT). A boiling water reactor (BWR) pin-cell problem under different power conditions is calculated and analyzed. The numerical results reveal that the proposed hybrid method improves the sensitivity-coefficients of eigenvalue and few-group homogenized cross sections. The temperature effects on the sensitivity-coefficients are demonstrated and the uncertainties are analyzed.     

Impact of the mass isotope on plasma confinement and transport properties in the HL-2A tokamak

The impact of the mass isotope on plasma confinement and transport properties has been investigated in Ohmically-heated hydrogen and deuterium plasmas in the HL-2A tokamak. Experimental results show that under similar discharge parameters the deuterium plasma has better confinement and lower turbulent transport than the hydrogen one, and concomitantly, it is found that the magnitude of geodesic acoustic mode zonal flows, the tilting angle of the Reynolds stress tensor and the turbulence correlation lengths are all larger in the edge region of the deuterium plasma. The results provide direct experimental evidence on the importance of the nonlinear energy coupling between ambient turbulence and zonal flows for governing the isotope effects in fusion plasmas.     

一个基于虚拟技术的幅度谱模拟与处理系统

概述了利用Visual C^ 6．0 Windows设计的一个幅度谱的模拟与处理系统。幅度谱的模拟包括：本底分布、峰分布和谱分布,而谱分析处理包括：谱平滑、寻峰、峰判别、扣本底、峰面积、能量刻度和效率刻度等。     

基于小波包多尺度信息熵和HMM的气液两相流流型识别方法

为了研究垂直上升管中气液两相流的流型,利用自制的多电导探针测量系统采集了四种典型流型的电导波动信息.由于气液两相流电导波动信号的非平稳特征,提出了一种基于小波包多尺度信息熵(Wavelet Packet Multi-scal2e Information Entropy)和隐马尔可夫模型(Hidden Markov Model,HMM)的流型识别方法.该方法首先对采集到的电导波动信号进行3层小波包分解,得到了8个不同频带的信号,提取各频带信号的小波包多尺度信息熵特征作为流型的特征向量,然后将其转换为观测序列输入到各种状态的隐马尔可夫模型进行训练并识别流型.结果表明:与BP神经网络相比,采用隐马尔可夫模型进行流型识别可以获得更高的识别率,表明该方法是有效和可行的.     

A method for 3D simulation of internal gas effects on thermal-mechanical behaviors in nuclear fuel elements

A new method for three-dimensional simulation of the interaction between the gas and the solid around is developed.The effects of the gas on the thermal-mechanical behaviors within the surrounded solid are performed by replacing the internal gas with an equivalent solid in the modeling,which can make it convenient to simulate the thermal-mechanical coupling effects in the solid research objects with gases in them.The applied thermal expansion coefficient,Young’s modulus and Poisson’s ratio of the equivalent solid material are derived.A series of tests have been conducted;and the proposed equivalent solid method to simulate the gas effects is validated.     

Seismic probabilistic risk analysis based on stochastic simulation of accelerograms for nuclear power plants in the UK

This article presents an approach to probabilistically assess the seismic risk of nuclear power plants (NPPs) in the UK. The approach proposed is based on direct stochastic simulation of the seismic input to conduct nonlinear dynamic analysis of a structural model of the NPP analysed. Therefore, it does not require the use of ground motion prediction equations and scaling/matching procedures to define suitable accelerograms as is done in conventional approaches. Additionally, as the structural response is directly calculated, it does not require the use of Monte Carlo-type algorithms to simulate the damage state of the NPP analysed. However, it demands longer use of computer resources as a relatively large number of nonlinear dynamic analyses are needed to perform. The approach is illustrated using an example of a 1000 MW Pressurised Water Reactor building located in a representative UK nuclear site. A comparison of risk assessment is made between the conventional and proposed approaches. Results obtained are reasonable and well constrained by conventional procedures; hence, it can confidently be used by the UK New Build Programme in the next two decades to generate 16 GWe of new nuclear capacity.     

国际原子能机构放射性物质运输安全标准发展回顾

国际原子能机构(IAEA)有5个安全标准委员分别负责相关领域安全标准制修订工作,运输安全标准委员会(TRANSSC)是其中之一,主要负责与放射性物质运输安全相关标准的制修订工作。文章介绍了IAEA放射性物质运输安全标准发展、TRANSSC的简要情况、标准制修订流程以及近年来SSR-6等文件修订情况,提出了我国参与IAEA运输安全标准制修订工作的建议。     

A Method for the Numerical Integration of the Photon Transport Equation in Slab Geometry

A procedure for the direct numerical integration of the transport equation in slab geometry is introduced. The angular flux is expressed as a series of Legendre polynomials. The iteration process is utilized for the spatial integration of the transport equation. Prior to applying this method both to neutrons and to photons, a series of studies on the photon penetration problem has been performed to assess its validity. The first code, the EOS-1, was prepared for the NEAC 2206 computer to perform the above mentioned work and also to solve some practical problems of not very large scale.

A comparison of energy buildup factors obtained by the EOS-1 code with those from the moments method and from the Monte Carlo method indicates that the method presented in this paper provides a correct treatment of the attenuation of radiations in a slab. The results also agree well with the energy spectra from the moments method, with the energy spectrum measured in the Bulk Shielding Reactor and with the energy-angular flux distributions obtained with ⁶⁰Co and ¹³⁷Cs sources behind iron and aluminum plates, both with respect to spectrum pattern and to absolute values.