核反应堆压力容器的基本设计准则讨论

目前,核动力反应堆的压力容器主要按照ASME《锅炉和压力容器规范》第Ⅲ卷“核动力装置部件”来设计。本文详细分析了这部规范所采用的各项基本设计准则的理论依据,并指出了使用这些设计准则应注意的事项。由于这部规范对应力作了合理的分类,并对不同的应力规定了不同的设计准则,同时又广泛地采用了塑性失效和塑性分析的方法,所以能在较保守的安全系数下有效地提高部件的设计工作应力,达到既节约材料又确保安全可靠的目的。 规范所提出的各项基本设计准则,除疲劳以外,主要是防止过度的塑性变形,所以可统称为防止塑性失效分析。它和核容器的防止脆性破坏分析和防止疲劳分析一起,成为核容器防止破坏的安全设计分析的三项基本内容。     

VENUS程序应力结果线性化处理程序V2MB的研发

在反应堆压力容器与锅炉的设计中，国际上比较通用的规范（包括美国的ASME规范、法国的RCC—M规范、日本的JIS规范等）在对设备进行评价时均要求对应力结果进行分类，将应力结果沿壁厚线性化处理，给出薄膜、薄膜加弯曲应力，然后在此基础上完成强度评价。     

反应堆压力容器强度可靠性分析

应用ANSYS有限元程序,采用蒙特卡洛法中的直接抽样法和拉丁方抽样法、响应面法中的中心指数设计抽样法和Box-Behnken矩阵抽样法完成反应堆压力容器强度可靠性分析,给出指定输入条件下压力容器强度的可靠度。结果表明,对压力容器母材可靠度的影响程度由大到小依次为内压、母材许用应力和母材弹性模量;对主螺栓可靠度的影响程度由大到小依次为螺栓材料许用应力、螺栓预紧力和内压。     

分析设计中应力分类的一次结构法

我国新的设计规范ＪＢ－４７３２－９５《钢制压力容器－分析设计标准》于１９９５年３月颁布实施。如何将有限元分析呈实验应力分析得到的总应力场分解在规范中定义的各种应力类别是应用ＪＢ－４７３２－９５或ＡＳＭＥ《锅炉及压力容器规范》等Ⅲ篇和第Ⅷ第２分岫时必须解决的关键问题。本文提出应力分类的两步法和一次地、将它们和等儿线性化方法相结合给出了圆满解决该问题的有效方法。文中还阐述了应力分解的不唯一性，自限应力     

论高强度钢在核动力反应堆压力容器中的应用

本文论述高强度钢在核动力反应堆压力容器中的应用前景.着重指出,在技术工艺水平高度发展的今天,限制高强度钢应用的主要障碍不在于材料和制造工艺,而在于现行的压力容器规范:建议改变现行规范规定的取用设计许用应力的方法,修改材料的极限强度对设计许用应力的限制,而代之以针对不同材料和不同类型的容器规定一个合适的材料屈强比要求.     

秦山核电二期工程反应堆压力容器设计

反应堆压力容器的设计、制造、安装和试验应与其安全功能相适应;采用公认法规和标准时,应对其进行评价,保证满足反应堆压力容器的安全功能.在设计、制造、安装和试验中,必须使异常泄漏、裂纹快速扩展及破坏的概率降低到最小.在初步设计阶段完成了强度设计后,进行了扩大设计状态的一次应力分析评定,结果均满足RCC-M-B册设计工况的各项应力准则.     

压力容器形状偏差对安全性能的影响

压力容器制造过程中总会存在形状偏差,设计者应能明确可接受的形状偏差是多少。本文以有限元应力计算为基础,分析了整体形状偏差所造成的筒体和封头的一次应力的变化情况及其对压力容器安全性能的影响,得出了一些具有普遍意义的结果,并以ＨＴＲ－１０反应堆压力容器为例,根据设计和制造中的具体情况,分析了可接受的形状偏差限值。     

高温气冷堆结构中的力学问题

本文着重介绍了高温气冷堆结构中的力学问题。由于高温气冷堆采用石墨作为主要结构材料,因而在高温、高辐照以及氧化气氛中石墨力学特性的研究是本文介绍的重点:诸如石墨在上述工作环境下的物理性质及力学性质,石墨的疲劳特性、石墨的应力分类及破坏准则等。对高温气冷堆压力容器(PCPV)应力分析中混凝土的徐变及开裂问题的研究,本文也做了介绍。     

反应堆压力容器快中子注量计算模型简化方法研究

反应堆压力容器快中子注量计算是反应堆工程设计和国家核安全监管部门关注的一项重要内容,在计算中不可避免地采用一些模型简化方法,这些简化方法将对计算结果产生一定的影响。利用MC方法计算程序MCNP对反应堆压力容器快中子注量计算过程中的模型简化方法开展研究,评估这些简化方法的利弊,为压力容器快中子注量计算提供合理建议,保障计算的严谨性。     

CRDM中部?密封环堆焊修复裂纹扩展分析

核电厂控制棒驱动机构(CRDM)耐压壳采用?密封环焊接安装在反应堆压力容器顶盖的管座上。一回路水应力腐蚀易诱发?密封环焊缝产生裂纹导致泄漏,需采用堆焊技术进行修复以保证?密封环结构完整性。基于ASME规范中的断裂力学分析方法,针对?密封环堆焊修复的设计结构,开展疲劳和应力腐蚀引起的裂纹扩展分析计算,为?密封环堆焊设计和评定提供参考。     

Products of research and development in NDT

The Nondestructive Ultrasonic Testing and Evaluation is directed to describe flaws microstructure, structure residual stress, reliability or quality insurance. An overview is given on the scientific background, interpretation capability based upon mathematic-numerical models, new design of equipments and demonstrated at components containing relevant flaws. Emphasis is laid upon SNR-improvement in the UT of austenitic weldings. Examples are shown using the backscattering technique with linear polarized shear waves. Stress measurements are performed with electromagnetic ultrasonic probes (EMAT). The surface inspection with magnetic imaging of the eddy current field or the multi-frequency eddy current methods are described. The possibilities of detection, classification and reconstruction of flaws are demonstrated in the time- and frequency domain with the satellite pulse echo technique, focal probes and the synthetic aperture methods like ALOK, Phased Array, Holography and SAFT.     

核电主设备分析法设计中应力线性化路径可靠性及优化分析

分析法设计是核电主设备设计的主要方法之一。该方法将结构设计或评定中各输入参量进行偏于安全的假设,以安全-不安全定性反映主设备设计的结构完整性状态。在确定性分析法设计的基础上,本研究基于概率统计理论,利用可靠性及优化分析方法,综合考虑结构设计或评定中涉及的主观不确定性因素(应力线性化路径选取),选取核电主设备典型结构形式——蒸汽发生器过渡锥体段为研究案例,首先对影响应力线性化路径选取的参数进行敏感性分析。随后,优化出最佳应力线性化路径。本研究方法为工程设计中应力线性化路径的选取提供建议。同时,该方法对可靠性理论在ASME核电规范与标准的分析法设计中的应用具有积极意义。     

Enhanced seismic criteria for piping

In situ or laboratory experiments have shown that piping systems exhibit satisfactory seismic behavior. Seismic motion is not severe enough to significantly damage piping systems unless large differential motions of anchorage are imposed. Nevertheless, present design criteria for piping are very severe and require a large number of supports, which creates overly rigid piping systems. CEA, in collaboration with EDF, FRAMATOME and IRSN, has launched a large R&D program on enhanced design methods which will be less severe, but still conservative, and compatible with defect justification during operation. This paper presents the background of the R&D work on this matter, and CEA proposed equations.Our approach is based on the difference between the real behavior (or the best estimated computed one) with the one supposed by codified methods. Codified criteria are applied on an elastically calculated behavior that can be significantly different from the real one: the effect of plasticity may be very meaningful, even with low incursion in the plastic domain. Moreover, and particularly in piping systems, the elastic follow-up effect affects stress distribution for both seismic and thermal loads.For seismic load, we have proposed to modify the elastic moment limitation, based on the interpretation of experimental results on piping systems. The methods have been validated on more industrial cases, and some of the consequences of the changes have been studied: modification of the drawings and of the number of supports, global displacements, forces in the supports, stability of potential defects, etc.The basic aim of the studies undertaken is to make a decision on the stress classification problem, one that is not limited to seismic induced stresses, and to propose simplified methods for its solution.     

一级设备的应力与应变分析研究

一级设备在核设备中非常重要,在该设备的设计和调试的过程中需要对其进行应力与应变分析,从而确保其安全性。对一级设备的应力与应变分析有两种方法,理论方法和实验方法。通过理论方法,可以得到一级设备在各种工况下的应力与应变状况,从而为设备的设计提供依据;通过实验方法,可以验证理论方法的准确性,并为设备的安全评定提供依据。中国实验快堆(CEFR)主容器属于一级设备,本文结合CEFR主容器的应力与应变分析研究,根据AMSE规范,描述了核安全一级设备的应力与应变分析的理论方法,同时也介绍了测量核安全一级设备的应力与应变的实验方法。     

Temperature and stresses in reactor fuel elements due to time- and space-dependent heat-transfer coefficients

In this paper various methods are discussed and compared for the prediction of temperature and thermoelastic stress fields in nuclear reactor fuel elements and other components due to both steady and transient non-uniform surface heat transfer. A general mathematical outline of the problem is given, with the sphere and infinite half-space of constant properties considered in detail. For a number of heat-transfer distributions results for temperatures and stress components obtained by various methods are presented. It is concluded that significant temperature and stress increases due to non-uniform heat-transfer variation must be taken into account in the design of nuclear reactor components.     

Evaluation of nuclear system requirements for accommodating seismic effects

An integrated approach is presented for the design of nuclear facilities to accommodate seismic effects. Site evaluation and soil-structure interaction are discussed briefly since they determine the magnitude and characteristics of the input forcing function to be used in the analysis. The evaluation of the requirements of the nuclear system to accommodate the effects of specified seismic input forces involves the classification of components of the system, development of a mathematical model which adequately defines and relates the components of the system, consideration of coupling and resonance effects on the interactions among the components, selection of the methods of analysis and choice of the methods of solving the resulting equations. Specific design considerations and criteria for judging the acceptability of the design are discussed.     

Shape optimization of a perforated pressure vessel cover under linearized stress constraints

One of the general methods to evaluate a failure condition is to compare a maximum stress with an allowable stress. A failure condition for a stress is usually applied to a concerned point rather than a concerned section. In an optimization procedure, these stress conditions are applied as constraints. But the ASME code that prescribes its general rules upon the design of a NSSS (nuclear steam supply system) has quite a different view on a failure condition. According to the ASME code Sec. III, a stress linearization should be performed to evaluate a failure condition of a structure. Since a few programs provide a procedure for a stress linearization through a post-processing stage, an extra calculation of the linearized stresses and the derivatives of a linearized stress are conducted to adopt the stress linearization results to an optimization procedure as constraints. In this research, an optimization technique that utilizes the results of a stress linearization as a constraint is proposed. The proposed method was applied to the shape design of a perforated pressure vessel cover.     

Advanced applications of boundary-integral equation methods

Numerical analysis has become the basic tool for both design and research problems in solid mechanics. The need for accuracy and detail, plus the availability of the high speed computer has led to the development of many new modeling methods ranging from general purpose structural analysis finite element programs to special purpose research programs. The boundary-integral equation (BIE) method is based on classical mathematical techniques but is finding new life as a basic stress analysis tool for engineering applications. This paper summarizes some advanced elastic applications of fracture mechanics and three-dimensional stress analysis, while referencing some of the much broader developmental effort. Future emphasis is needed to exploit the BIE method in conjunction with other techniques such as the finite element method through the creation of hybrid stress analysis methods.     

核电厂核2级承压管道抗震设计规范对比分析

RCC-M、ASME(2007版)及GB 50267-97为目前核电厂设备、系统、部件设计所遵循的主要技术标准,3者对核电厂部件的分级基本相当,在核2级承压管道设计方面的规定内容相似但不完全相同.在地震输入方法上,GB 50267-97、ASME(2007版)及RCC-M基本相同,GB 50267-97中硬土场地的水平...     

Multiaxial fatigue analysis for IMIC of ITER upper ELM coil

Inconel Jacketed Mineral Insulated Conductor (IMIC) is a very important component of International Thermonuclear Experimental Reactor (ITER) Edge Localized Modes (ELM) coils, which are located between the vacuum vessel (VV) and blanket shield modules and subject to high radiation levels, high temperature and high magnetic field. These coils will experience thermal pulsed, cyclic electromagnetic (EM) load during operation. They are designed to sustain at 1.5e8 total stress cycles and shall have sufficient strength and excellent fatigue to transport and bear the high cyclic load. For IMIC, multiaxial fatigue analysis is used to evaluate failure. Two methods based on the alternating stress and mean stress in American Society of Mechanical Engineers (ASME) code provide the design codes for multiaxial fatigue evaluation: constant principal stress direction and variation of principal stress direction. Results show that using the two methods obtains basically the same equivalent alternating stress. Both of them can be recommended for the ELM coils and IMIC can meet the fatigue criteria.