用样条函数和统计技术作拟合曲线

本文介绍用样条函数和统计技术作拟合曲线的方法。简单讨论了三次B-样条及其主要性质,给出了一般样条加权拟合公式,以及除对各实验值分配权重外,对各家实验数据作为一个整体也分配适当权重的两次应用正态分布给出的样条加权拟合公式。最后讨论了程序的使用情况和存在问题。     

Ge(Li)γ谱仪的效率刻度

在60～3500keV能区,用~(75)Se,~(152)Eu及~(56)Co对Ge(Li)探测器作了相对效率刻度,并用标定过强度的~(113m)In、~(60)Co作了绝对效率刻度。应用三次样条函数做拟合内插,由加权平均的归一化系数,给出这能区中33个点的绝对效率及其误差值。γ能谱由自动解谱程序分析。能区中任一点的绝对效率及其误差,可由三组三次样条函数拟合内插程序算出,绝对效率值的误差<±4％。     

CNCSES-计算机化的核截面评价系统

CNCSES是人机会话的核截面评价系统,利用它可以对EXFOR实验数据和理论计算数据评价给出成套的核截面数据。系统有下列部分组成:EXFOR数据检索,格式转换,数据的分析比较,曲线拟合,多条曲线的统调,评价数据格式的标准化和检查。文中重点叙述了利用Bayes方法的单能点数据处理和在相同能点的统调,多家数据样条拟合的节点优化和拟合误差的严格计算以及用B样条函数对多条曲线的统调拟合。     

任意形状束流相空间变换的样条函数表述

为了研究具有任意形状相空间的束流的传输,本文提出了束流相空间变换的样条函数表述。利用三次样条函数拟合相空间边界,将相空间变换化为样条函数的变换,并进而计算束流包络。给出了具有任意形状相空间的束流传输的实例。     

正电子湮没寿命谱指数函数拟合程序

本文提供一个从正电子湮没寿命谱中消除时间分辨函数的影响作指数函数拟合的方法。用剥皮法给出参数初值和拟合数据的起始点,使所用高斯-牛顿法中的迭代过程迅速而准确地达到收敛。编制的计算程序(PLSEF)经使用证明效果是好的。     

利用样条函数进行核截面的统调拟合

一、引言在核数据的评价中,为了给出成套的、自洽的核数据,在分別独立完成的各个反应道的评价和有关的理论计算的基础上进行统一的调整是必须的,为此我们发展了利用样条函数进行统调拟合的方法。     

堆芯轴向功率分布曲线拟合

本文简要地叙述了用 INCA 法将固定的堆内自给能探测器的“读数”拟合成轴向功率分布曲线,提出了将固定的有限长度的自给能探测器的“读数”经简化处理后就可将拟合函数为 x 的4次多项式的最小二乘法进行曲线拟合。选择典型的功率分布曲线,用 INCA 法、拟合函数为 x 的4次多项式最小二乘法和二次样条函数法处理,并比较其结果。考虑了格架的影响后,用拟合函数为 x 的4次多项式的最小二乘法得到的结果是好的,它的计算比样条函数法简单。用这方法还可省去长钒自给能探测器。     

SPAS-G:高分辨率γ谱定量分析程序

本文叙述了一个高能量分辨率γ谱定量分析程序。由于程序中包含了几种不同的峰分析方法,对不同的谱数据和不同的计算机系统都有良好的适应能力。文中对于适应于微型计算机的线性函数拟合计算峰净面积方法的精度也进行了讨论。     

反应率相对分布实验数据处理程序的研制

为处理CEFR反应率相对分布实验准备数据，研制了活化法测量反应率相对分布实验数据处理程序。步骤如下：1）推导了活化法测量相对反应率计算表达式，引入功率归一因子，以便将径向各测点的照射功率归一到相同水平；2）推导了热堆轴向、径向中子注量率相对分布曲率拟合计算公式：3）推导了快堆中子注量率分布三次样条函数拟合计算表达式。     

正电子寿命谱测量中双功能程序PLRF的设计

在正电子寿命谱测量中,要处理的数据有两类——分辨函数及寿命谱数据。由于二者同属非线性函数的拟合问题,故可归于同一程序中解决。本程序就是利用最小二乘原理观点而设计的。处理分辨函数是先用二次曲线拟合原始数据,而后用单高斯线型对原始数据进一步拟合,从而得到仪器的分辨率等重要参数。处理寿命谱是按多指数拟合。其初值的给出有两种形式,一种是由程序用剥皮法求得初值;另一种是人为输入初值,并可根据计算过程偏差的变化,方便地对输入初值进行调整。     

多家相关数据的样条拟合

多家数据的样条拟合，已有不少工作，但都没考虑相关误差。文章处理了相关数据，推导公式并编写程序，研究了相关性对拟合值的影响，对ＰＰＰ问题也作了讨论。     

实验数据同时评价的样条函数方法

研究了比率相关同时评价问题。利用２种实验数据（几种截面及其相应的比值）来求截面的拟合值，并给出误差估计。将截面值的对数用样条函数来逼近。按广义最小二乘法推导计算公式，并介绍、分析算例。     

Comparison of three applications of ConTraSPECT

Today, the trend in SPECT attenuation correction is to acquire a transmission data set and reconstruct the corresponding attenuation map, for incorporating into the emission reconstruction. Recently, some techniques have been developed to determine the attenuation map directly from the emission data, by exploiting the fact that in the absence of noise only some specific transmission data can be consistent with given emission data. However it is necessary to include some a priori information so that the consistency conditions can be used in an efficient manner. Here, three types of priors are used: a uniform ellipse, a uniform spline curve, and a uniform spline curve knowing the untruncated part of the transmission sinogram. Once the corresponding attenuation map has been computed, it is used to correct for attenuation in the emission image reconstruction process. The three methods have been tested on both simulations and experimentally acquired data. The results show that the two first methods give equally acceptable results, while the third method provides reconstructions which are closer to reconstructions obtained when the attenuation map is computed from untruncated transmission data     

Comparison of stress-based and strain-based creep failure criteria for severe accident analysis

We conducted a parametic analysis of stress-based and strain-based creep failure criteria to determine if there is a significant difference between the two criteria for SA533B vessel steel under severe accident conditions. Parametric variables include debris composition, system pressure, and creep strain histories derived from different testing programs and mathematically fit, with and without tertiary creep. Results indicate significant differences between the two criteria. Stress gradient plays an important role in determining which criterion will predict failure first. Creep failure was not very sensitive to different creep strain histories, except near the transition temperature of the vessel steel (900 K to 1000 K). Statistical analyses of creep failure data of four independent sources indicate that these data may be pooled, with a spline point at 1000 K. We found the Manson-Haferd parameter to have better failure predictive capability than the Larson-Miller parameter for the data studied.     

非平面对称磁场的空间展开及样条函数表述

为了研究带电粒子光学系统的光学特性,需要知道系统中磁场的空间分布。实际磁场往往不具有严格的平面对称性质,一些作者曾就这类磁场的空间展开做过不少工作。但他们的工作是在柱坐标系统中进行的,适用于诸如回旋加速器、双向聚焦磁分析器、环形磁β谱仪等一类装置的磁场分析。对于大型同位素电磁分离器(calutron)那样一类装置,其磁场分布需要     

Uncertainty quantification in (α,n) neutron source calculations for an oxide matrix

We present a methodology to propagate nuclear data covariance information in neutron source calculations from (α,n) reactions. The approach is applied to estimate the uncertainty in the neutron generation rates for uranium oxide fuel types due to uncertainties on 1) ^17,18 O(α,n) reaction cross-sections and 2) uranium and oxygen stopping power cross sections.The procedure to generate reaction cross section covariance information is based on the Bayesian fitting method implemented in the R-matrix SAMMY code. The evaluation methodology uses the Reich-Moore approximation to fit the ^17,18 O(α,n) reaction cross-sections in order to derive a set of resonance parameters and a related covariance matrix that is then used to calculate the energy-dependent cross section covariance matrix. The stopping power cross sections and related covariance information for uranium and oxygen were obtained by the fit of stopping power data in the α-energy range of 1 keV up to 12 MeV.Cross section perturbation factors based on the covariance information relative to the evaluated ^17,18 O(α,n) reaction cross sections, as well as uranium and oxygen stopping power cross sections, were used to generate a varied set of nuclear data libraries used in SOURCES4C and ORIGEN for inventory and source term calculations. The set of randomly perturbed output (α,n) source responses, provide the mean values and standard deviations of the calculated responses reflecting the uncertainties in nuclear data used in the calculations. The results and related uncertainties are compared with experiment thick target (α,n) yields for uranium oxide.     

一种数据归档机制的实现

有效而完整地及时归档大型实时处理系统的原始数据是一项十分重要的工作.论文讨论的数据归档机制能每天自动地存储和归档国家数据中心所有要归档的数据,同时保证用户能有效检索已存储到磁盘阵列上的数据,在确认数据正确归档后,自动清除运行文件系统中的数据文件和运行数据库表中相关记录.该机制已在国家数据中心稳定运行.     

评价寿命统计分布的信息量模拟

提出了有限数据下寿命分布统计方法的综合评价模型。在确定寿命分布时,主要考虑总体拟合效果、与疲劳物理的一致性和尾部预测安全性3个因素：用失效概率试验值与理论值的相关系数评价总体拟合效果;与疲劳物理的一致性按照给定应力下失效率是否随循环数的增加而增加来确定;通过识别统计分布尾部的变化趋势来完成安全性的预测。通过引入信息理论．在寿命统计分布满足以上3个因素的前提下,建立了综合评价统计分布优劣的定量模型。通过实例分析．阐述了方法的具体应用。     

D-optimal design of fission-track annealing experiments

Fission tracks have been used to estimate time–temperature (t–T) paths of mineral samples of geological interest. Several datasets on annealing rate have been presented, especially for apatite. However, there are still open problems related mainly to chemical composition of apatite and crystallographic orientation of the analyzed crystal sections. Therefore, new annealing data are desirable. However, there are practical limitations (cost and mainly time) on the number of experiments that can be carried out. In this work, a method for optimizing the choice of the experimental conditions of the future experiments is presented, the so-called D-optimal criterion. The D-optimization takes advantage of previous knowledge on the experimental space and the model to be fit, to maximize the information to be obtained from new experiments. It has proved, in other instances, to be suitable for highly non-linear models as the Arrhenius-type equations currently used to fit the fission-track data. A sample example is given, showing that this procedure produces good choices, leading to reliable data fitting.     

核数据评价与中国评价核数据库CENDL

核数据是核基础研究、核能开发与利用以及核技术发展的基础数据,是连接核物理基础研究与核工程和核技术应用的重要桥梁,在国防与国民经济建设以及核科学发展领域起重要作用。核数据评价建库与检验是核数据研究过程中的两个重要部分,是核数据应用于核工程必不可少的环节。本文介绍了核数据内涵、核数据研究意义以及国内外核数据评价研究的简要发展历史,并结合中国评价核数据库CENDL的研究过程介绍了实验数据调研与分析评价、核数据理论模型计算、核数据统调建库与核数据宏观检验的主要评价核数据研究过程,以及我国自主建立的核数据评价方法和技术、模型及计算程序、评价数据建库和评价数据库的检验方法;介绍了基于我国自主建立的核数据评价建库与检验系统而研制的中国评价核数据库最新版CENDL 32以及对其进行的相关基准检验及应用结果;最后简要介绍了CENDL 32在反应堆屏蔽设计以及压水堆、高温堆等方面的实际应用以及与其他主流评价核数据库的比对结果。