针对一种简易链长测量法的误差分析及改进方案

 目前在链磨损试验中用于检验延伸率的一种简易链长测量方法在其计算过程中存在较大理论误差,面对日益多样化的测试任务不具有普遍有效性,有必要对其改进．将基于中心距测量的测量系统拆分成与中心距对应的绷直链部分和与测量设备接触的弯折链部分,对测量系统进行数学建模,同时考虑具体的测量时刻对测量误差的影响,对误差展开较为全面的分析,结果发现误差主要来源于用弧长代替折线的近似手段．随后提出的一种基于理论节距的改进算法经过理论分析和对照测试标准的计算证明,在整体误差上比原算法小1个数量级,能显著地提高简易测量法的精度以满足工程要求,拓展了其有效应用领域．此外,结合设备的误差分析还以提高精度为目标对测试工具的设计提出了建议．     

利用计量型原子力显微镜进行纳米台阶高度测量

计量型原子力显微镜纳米测量系统主要由扫描器、测针位置传感器和一体化微型激光干涉三维测量系统等部分构成．针对计量型原子力显微测量系统,采用三维激光干涉测量系统作为测量基准,以实现原子力测量系统的纳米尺度量值溯源和校准工作．建立了校准模型,分析了扫描器9项主要误差项,并将该模型应用到原子力显微镜扫描器的校准中．校准后的结果表明,除z轴位置误差不超过±2nm外,其他8项的残余误差均不超过±1nm．通过台阶高度国际比对,建立了台阶高度标准计算方法及不确定度分析模型．台阶高度国际比对的测量结果表明,计量型原子力显微镜的测量值与参考值相差均小于1．5nm．     

某球面对比测量系统误差分析模型研究

结合某球面对比测量系统的结构特点,建立了该系统的误差分析模型,采用微分方法计算分析了该系统测量过程中的误差变化规律,通过算例验证了该误差计算结果,据此给出了该测量系统的误差影响因素。     

微小深度尺寸测量系统的误差与测量不确定度分析

该文针对"微小深度尺寸现场测量系统的研究"项目进行原理阐述和误差来源的深入分析,同时对各类误差引入的测量不确定度进行计算,利用计算结果来验证该测量系统的测量不确定度。     

人工神经网络蒸汽流量计量误差校正方法

针对影响蒸汽流量测量精度的因素较为复杂,难以对测量误差进行精确的理论计算的特点,采用人工神经网络误差校正模型去除流量测量系统误差的方法来提高测量精度。在不改变原有测量装置的条件下,提高了流量测量系统的精度,校正后测量误差较原测量系统减小约7．8％。     

船用柴油机气缸套几何误差在线测量技术研究

设计一套船用柴油机气缸套几何误差在线测量系统,分析机床定位误差、传感器安装及测量误差对缸套表面各被测点坐标计算准确度的影响,建立测量误差计算数学模型。基于非线性规划方法建立缸套直线度误差、圆度误差、圆柱度误差的评定数学模型,通过与现有测量数据及误差评定模型的对比分析,表明该模型评定准确度较高。搭建在线测量原型系统,并进行测量实验,实验数据表明该系统具有结构简单、检测准确度及检测效率高等特点。     

基于标准球的显微图像测量系统标定方法

提出一种利用标准球对显微图像测量系统进行标定的方法．该方法利用显微系统小视场小景深的成像特点简化计算模型,减少标定参数．通过标准球图像在水平和垂直方向上的直径比计算比例因子;利用标准球边缘图像的边缘点集．运用优化的方法来计算成像系统的畸变系数和主点位置．系统的放大倍数由标准球的实际直径来标定得出．利用标准球在多摄像机公共视场内其轮廓在任何位置均可见这一特性,可同时对显微图像测量系统中的多个摄像机进行标定,简化标定过程．实验结果表明,该方法标定精度较高,标定后的测量系统的极限误差3σ为2．4μm．     

配网高压电能计量装置整体校准技术研究

该文对高压电能计量装置直接测量系统的运作原理和结构进行了介绍,并且运用这套系统直接测量10kV、100 A的高压电能计量装置的整体误差,将测量结果与计算过的综合误差法的结果进行对比,结果表明高压电能计量装置的直接测量系统对高压电能计量装置的整体误差可以进行非常准确的测量,比需要计算的综合误差法更加快速、直接、准确,高压电能计量装置直接测量系统的运用可以有效提高在高压电能计量装置的实际运行过程中其误差性能的稳定性。     

标量网络分析仪误差校正分析及其应用

为了消除标量网络分析仪在测量过程中的系统误差．提高测量精度,采用误差分析及校正模型．用数值分析的方法去除系统误差,降低随机误差．理论计算和实际测量结果都表明,测量精度明显提高．该误差分析理论及误差测量方法在标量网络分析仪中得到了较好的应用．提高了仪器的性能．     

衍射法X射线激光分束光栅平行度测量系统(英文)

介绍了软X射线光栅分束镜Mach-Zehnder干涉等离子体诊断系统调整用双频光栅线条平行度的衍射法测量系统.该系统主要由激光器,准直镜,待测光栅,精密转台(含角度测量仪),直线工作台,光栅调节架和探测器组成.分析了系统中各种误差对测量精度的影响其中包括距离测量误差、波长误差、光栅准直误差即光栅刻线与入射光和反射光组成的平面不垂直、光栅转动过程的误差、光栅表面面型误差、探测器误差,经计算得到系统的绝对误差为minute.计算表明,该系统的测量精度满足软X射线Mach-Zehnder干涉系统对双频调整光栅的性能要求.     

加热式热电偶液位测量传感器的研究

压力容器内的水位是反应堆运行中的重要参数,针对传统的传感器难以实现液位测量,提出了一种加热式热电偶传感器, 它是基于发热体在气(汽)体和液体中放热系数的显著差异,研制的一种变液位测量为温差测量的加热式铠装热电偶.试验表明,液位测量传感器原理正确,结构可行,性能可靠,能够准确的判断出液气(汽)界面的位置,推广该传感器在压力容器内应用很有必要.     

高压加热器最佳运行水位的确定方法研究

分析了加热器水位变化对上下端差的影响,并采用等效焓降法建立了加热器上下端差对机组相对内效率影响的计算模型,提出了加热器最佳运行水位的确定方法。以某电厂600MW机组高压加热器为例,定量分析了正常低水位运行对机组热经济性的影响,并利用本文提出的方法对三台高压加热器水位进行了调整,调整后机组相对内效率提高0.070%,热耗率降低了5.20kJ/（kW·h）,该方法能推广到其他类似机组上应用,对电厂经济运行具有很大实用价值。     

Gamma dose from activation of internal shields in IRIS reactor

The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressuriser and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield.     

EPR主控室空调系统设计

在核电厂运行和事故工况下,其主控室空调系统都必须运行以维持主控室的可居留性和设备运行的环境条件,主控室空调系统设计是核电站最具有代表性的暖通系统设计.本文从欧洲压水堆（EPR）设计理念入手,对主控室空调系统安全功能的实现、系统配置、可靠性、运行、重要设备选型以及噪声控制方面进行深入分析,并对系统设计不足提出改进意见.     

Stress-Strain State Evaluation of a Welded Joint of Hot Collector to Nozzle of NPP Steam Generator PGV-1000

The paper describes a method and main results of a refined stress-strain state evaluation for a welded joint of the hot collector to steam generator nozzle of a nuclear power plant with a WWÉR-1000 reactor, which takes into account a fuller range of the actual loading conditions in operation. In calculations, we considered the stresses induced by uncompensated thermal expansion of the reactor elements. The results of 3D finite element calculations suggest that the level of local stresses in the area of the joint exceeds the metal yield stress. The calculated stresses for the outer nozzle surface agree satisfactorily with the in-situ strain measurement data.     

Performance of a drag transducer under simulated PWR refill/refloodflow conditions

A strain gage-based drag transducer developed to measure two-phase flow in simulated PWR (pressurized-water reactor) test facilities is discussed. The drag sensor was designed to sample a large amount of the flow with minimum disturbance to the flow, to be sensitive over a 1000-to-1 range, to be accurate over a 200:1 turndown ratio, and to survive in a hostile steam/water environment. The drag transducer consists of a unique combination of an elongated proving ring, a temperature-compensated, full-bridge strain gage configuration, and a drag target that utilized part of the test vessel internals. The drag body measurement system produced high-quality data for a variety of test conditions, which has led to a better understanding of the phenomena occurring during the refill/reflood stages of a postulated PWR loss-of-coolant accident     

浅谈锅炉汽包差压式水位计的维护

锅炉汽包水位测量的重性是人所共知的,汽包水位测量仪表的合理应用是保证锅炉安全稳定运行的重环节。文章结合吉林通化钢铁股份有限公司动力厂的实际情况,以锅炉汽包差压式液位计为例,分析其工作原理及在实际维护中的注意事项和存在问题,供仪表维护人员参考。     

Micromechanical modelling of reactor pressure vessel steel

Safe operation of nuclear power plants is one of the major concerns and relies primarily on the integrity of the reactor pressure vessel. Neutron exposure induces temperature dependent embrittlement and can lead to loss of fracture toughness of reactor pressure vessel steel. Conventional fracture mechanics suffers from a serious limitation of non-transferability of fracture data from specimen level to component level as fracture resistance data obtained is largely geometry dependant. This difficulty is largely overcome by Gurson–Tvegaard–Needleman (GTN) model which models the drop in load carrying capacity of a material with the increase in plastic strain, considering nucleation, growth and coalescence of micro-voids in the material. However, determination of the model-parameters with experiments is extremely difficult and doubtful. Hence, a parametric study was undertaken to find out the effects of different parameters on material behaviour of reactor pressure vessel steel and to asses the parameter for best practical result with least metallographic study using a hybrid approach. The model parameters were finally verified by predicting ductile fracture in compact tension and three point bend specimen made from reactor pressure vessel steel.     

分散液液微萃取浮动溶剂固化——原子吸收光谱法测定环境水体样品中的重金属

采用了十一醇为萃取剂,吡咯烷二硫代甲酸铵(APDC)为络合剂的分散液液微萃取浮动溶剂固化-石墨炉原子吸收光谱法(DLLME-SFO-GFAAS)测定水体样品中痕量金属离子的分析方法,对实验条件进行了优化,改进了传统萃取方法中试剂用量大、手续繁琐、易交叉感染等不足。同时对方法测量检出限及回收率进行试验,结果表明在优化的实验条件下铜、镉、铅的检出限分别0.03μg/L、0.03μg/L和0.01μg/L,加标回收率在95.3%--102.4%,方法准确、可靠。     

A decision theoretic approach to an accident sequence: when feedwater and auxiliary feedwater fail in a nuclear power plant

This study applies a decision theoretic perspective on a severe accident management sequence in a processing industry. The sequence contains loss of feedwater and auxiliary feedwater in a boiling water nuclear reactor (BWR), which necessitates manual depressurization of the reactor pressure vessel to enable low pressure cooling of the core. The sequence is fast and is a major contributor to core damage in probabilistic risk analyses (PRAs) of this kind of plant. The management of the sequence also includes important, difficult and fast human decision making. The decision theoretic perspective, which is applied to a Swedish ABB-type reactor, stresses the roles played by uncertainties about plant state, consequences of different actions and goals during the management of a severe accident sequence. Based on a theoretical analysis and empirical simulator data the human error probabilities in the PRA for the plant are considered to be too small. Recommendations for how to improve safety are given and they include full automation of the sequence, improved operator training, and/or actions to assist the operators' decision making through reduction of uncertainties, for example, concerning water/steam level for sufficient cooling, time remaining before insufficient cooling level in the tank is reached and organizational cost-benefit evaluations of the events following a false alarm depressurization as well as the events following a successful depressurization at different points in time. Finally, it is pointed out that the approach exemplified in this study is applicable to any accident scenario which includes difficult human decision making with conflicting goals, uncertain information and with very serious consequences.