核安全一级主管道疲劳校核

本文对某核电厂主管道疲劳及热棘轮进行了独立校核。校核采用基于RCC-M标准的ROCOCO软件,比较了RCC-M标准与ASME标准在核安全一级管道疲劳评价方面的差异。对比的主要方面包括疲劳设计的计算范围界定、一次加二次应力强度的计算方法、弹塑性修正系数的计算、动态载荷叠加方法等。通过对ROCOCO中与ASME标准不一致的算法进行修正,得到主管道冷段壁厚65 mm和55mm的疲劳使用系数和热棘轮设计裕量。结果表明:某核电厂主管道最小壁厚不能小于55 mm,55mm壁厚的热棘轮设计值达到许用值的95%。     

先进压水堆核电厂主管道结构完整性的研究

以某先进压水堆核电厂主管道为例,对核安全一级管道的结构完整性进行分析评价,并对根据规范设计的管道设计裕量进行了分析。管道结构完整性评价内容包括依据规范对管道强度进行评价、采用解析法求解管道温度场进行热棘轮评价、采用简化雨流法对管道进行疲劳寿命评价。计算结果表明,主管道最小壁厚减少至55 mm能够满足标准规范要求,但安全裕度较小,其中主管道支管位置的疲劳和热棘轮评价结果裕量最小。     

单位瞬态法在核电厂疲劳监测系统中的应用

核电厂疲劳监测系统中的非稳态导热反演计算是关键步骤,本文提出的单位瞬态法目的就是解决反演计算问题。首先研究了一维瞬态热传导内外壁温度的线性关系,其次建立了单位瞬态热传导有限元模型,然后通过数值计算实现了由内壁温度计算外壁温度或由外壁温度计算内壁问题双向计算过程,最后通过1组试验数据验证单位瞬态法的正确性,为疲劳监测系统的推广应用奠定了基础。     

反应堆压力容器简体60年寿命疲劳裂纹扩展分析

利用ANSYS程序,对反应堆压力容器筒体在正常运行工况下进行疲劳裂纹扩展分析,获得了反应堆压力容器筒体在60年寿期末的疲劳裂纹尺寸,按照RCC-M规范的要求,对压力容器在主管道破裂瞬态和主蒸汽管破裂瞬态下进行了快速断裂评价。研究结果表明,压力容器满足RCC-M规范的要求,不会发生裂纹失稳。     

压水堆核电站不锈钢主管道铸造

用电弧炉和ＡＯＤ双联冶炼核电站主管道Ｚ３ＣＮ２０－０９Ｍ，并根据Ｓｈａｅｆｆｌｅｒ图计算结果调整Ｚ３ＣＮ２０－０９Ｍ的铁素体含量。在离心铸管工艺中，用加大型筒壁厚，减小挡枝内孔直径，选大的重力加速度ｇ值，增加内孔加工余量等措施铸造出主管道样件，测试结果表明，主管道样件各项性能指标均满足ＲＣＧ－Ｍ的要求。     

严重事故下反应堆压力容器下封头耦合烧蚀传热分析

核电站严重事故发生后,反应堆压力容器（RPV）固壁在熔池作用下会发生烧蚀、减薄。开展RPV下封头耦合烧蚀传热分析对堆坑注水有效性论证和RPV剩余壁厚确认有重要的理论指导意义。本文以CPR1000反应堆压力容器为研究对象,在FLUENT 17.2平台下,基于动态网格方法和UDF二次开发,构建了综合考虑RPV固壁瞬态烧蚀与导热、RPV内壁热流密度再分布及RPV外壁过冷沸腾的全耦合计算模型,获取了9 000 s内的堆坑两相流场分布和RPV固壁烧蚀温度场,分析确定了最小剩余壁厚和发生位置。结果表明：使用动态网格捕捉壁面烧蚀的方法可行,本文全耦合计算模型在分析RPV固壁瞬态烧蚀过程方面有一定优势。     

CPR1000核岛安装主管道自动焊精确组对技术

主管道的精确组对是主管道自动焊技术成功实施的前提。立足于各个主设备及主管道的几何特征,提出针对性的竣工尺寸测量方案,据此计算出主管道坡口加工数据,以期消化主设备的制造误差;然后通过制定坡口加工方案保证坡口加工精度,从而最终保证主管道安装精确组对。     

船用堆燃料棒包壳疲劳寿命分析

船用堆瞬态变工况下燃料棒包壳温度和冷却剂压力波动较大,引起包壳的疲劳损伤,因此包壳疲劳寿命分析至关重要。本文利用ANSYS软件模拟船用堆瞬态变工况下燃料棒的热机械行为,结合锆包壳疲劳寿命设计曲线,考察包壳温度、冷却剂压力、燃料棒内压以及辐照对船用堆燃料棒包壳疲劳寿命的影响。计算结果表明,瞬态变工况使得包壳疲劳寿命有很大降低;包壳温度变化与冷却剂压力变化相比,前者对包壳疲劳寿命的影响小;辐照会降低包壳疲劳寿命。在不影响核动力船舶机动性的前提下,可采取一些必要的措施来降低包壳的疲劳损伤。     

蒸汽发生器管板二次侧表面温度场瞬态计算

蒸汽发生器管板二次侧表面温度场瞬态分析用于得到瞬态工况下管板二次侧附近流体的温度场分布,为管板的疲劳断裂分析提供输入数据。通过对法国管板二次侧表面温度场瞬态分析软件MYRTE的研究,在掌握软件建模和分析方法基础上,采用FLUENT实现管板二次侧流体温度场的瞬态计算。通过在控制方程中添加附加的质量源项和能量源项建立了二次侧流场的计算模型。质量源项中添加通过流量分配板流失的质量;能量源项中添加通过流量分配板的焓通量和来自一次侧的释热。计算得到的二次侧流体区域速度场和温度场分布与MYRTE计算结果符合较好,二次侧流体温度从入口至管板中心处是缓慢升高的,管板中心处加热较为明显。同时绘制出二次侧流体温度随时间变化的曲线图,可以为管板的疲劳断裂分析提供输入参数。     

一体化装置实验对RELAP5的评价

针对一体化自然循环试验装置OSU-MASLWR开展的实验,采用系统分析程序RELAP5/MOD3.3进行分析计算。失水事故瞬态计算结果表明,堆芯有足够的冷却,加热元件在整个瞬态过程中温度不断降低。安全壳内出现热分层现象,通过安全壳壁从安全壳到周围水池的热传递速率足以除去堆芯的衰变热。与试验结果相比,程序基本预测了整个瞬态过程中各参数的变化情况。     

Piping research overview

The USNRC Piping Review Committee (PRC) was formed in 1983 with a charter to review NRC piping criteria, to recommend changes to this criteria, and to identify areas that would benefit from future research. This overview will outline the NRC-sponsored research being conducted to address those PRC recommendations concerning the design of nuclear piping systems to withstand dynamic loads. A key element of this research is the joint EPRI/NRC “Piping and Fitting Reliability Research Program.” This program consists of dynamic capacity testing of piping at the system, component, and specimen levels, plus analyses needed to support recommendations for changes to the ASME Code. As part of NRC's contribution to the EPRI/NRC program, a pipe system capacity test will be conducted at ETEC. The “Nonlinear Piping Response Prediction” project at HEDL is evaluating nonlinear response prediction techniques with differing degrees of complexity and will compare the various analytical results both with each other and with physical benchmarks such as the ETEC test. An ORNL project is developing nozzle design guidance that will provide a more realistic basis for evaluating the higher nozzle loads that will result from the more flexible piping systems design that are being considered. INEL will evaluate high frequency damping by considering the existing high frequency data and by conducting high frequency/high stress tests on two piping systems. LLNL is now conducting studies to more completely assess the uncertainties in the seismic response of building structures and piping systems. As a follow-on to the research efforts reported in NUREG/CR-3811, BNL will conduct additional studies to improve combinational procedures for piping response spectra analyses.     

压缩机复杂管路压力脉动及管道振动研究

围绕往复式压缩机管道系统的振动及往复式压缩机的管道压力脉动问题,依据平面波动理论,采用转移矩阵和刚度矩阵计算出复杂管路的气柱固有频率和压力脉动.借助于有限元方法的离散思想,建立了往复式压缩机管道振动及应力分析的数学模型,提出了恰当的边界条件,利用基于有限元的管道分析软件CAESARⅡ对模型进行求解,获得了管道系统的振动模态结果.对比试验结果与计算结果发现,利用一维平面波动方程可以比较准确地计算出往复式压缩机管路的气柱固有频率、压力脉动.     

Piping system buckling analysis

A piping system with long horizontal runs hung by rods with pinned connections may require evaluation of its structural buckling characteristics if the piping is subjected to sustained upward-lifting loads. Piping systems in offshore installations or above the suppression pool of BWRs are examples of such systems.The paper initially reviews the fundamental theories on structural buckling of single support models, then derives a governing equation for structural buckling of piping systems with multiple supports, and finally discusses the mathematical significance of the stability as an eigenvalue problem. Four stability criteria based on critical load, energy, critical load ratio, and displacement ratio are presented. Applications of those criteria to the solution of stability problems are suggested. A hand calculation solution to an illustrative example is given.     

Piping extreme dynamic response studies

Results are presented for a series of high-amplitude dynamic tests of a simple pressurized piping system excited through various multiple piping supports. The four-inch diameter piping achieved response levels above yield when subjected to earthquake-like time history inputs and withstood — without leakage or gross distortion — dynamic inputs that were factors of three to five times greater than those inputs required to just exceed the ASME Class 2 stress limit for Service Level D, the Safe Shutdown Earthquake condition. Despite intentionally induced support failures in several tests, piping pressure integrity was maintained, and no plastic collapse occurred. Selected snubber hardware likewise exhibited large design margins under transient loads.     

Piping fracture mechanics data base

A comprehensive, computerized data base of fracture toughness (J - R curve) and other support data from nuclear piping steels is being established. This data base will be used for the materials input to assessments of piping integrity with known or assumed flaw and loading conditions.This data base will be accessible to NRC personnel, contractors and other outside users, as required. Regular updates will be made as more data are generated at MEA and by others with close coordination of testing plans tied to other NRC-sponsored research programs     

核级管道计算程序的应用

开发一种核级管道计算程序,可采用多种规范对核级管道进行应力分析与评定。介绍程序计算原理,并以某核电工程管道系统为例,采用RCC-M和ASME规范进行计算,分别和SYSPIPE、PIPESTESS的计算结果进行对比。计算结果表明,开发的程序计算结果正确,精度满足要求。     

多级节流孔板在核级管道中的应用

针对大亚湾核电站安全壳喷淋系统(EAS)试验管线节流孔板气蚀引起的管道剧烈振动和噪音,以及支管疲劳破坏这一事例,研究了气蚀引起管道振动的分析方法,以及采用多级节流孔板减小气蚀的设计方法.对气蚀引起的管道振动,采用计算流体动力学(CFD)方法分析孔板附近的流动特性和压力分布,确定节流孔板下游是否发生气蚀现象;对于发生气蚀现象的节流孔板,提出采用多级节流孔板来减弱气蚀,并采用各级节流孔板气蚀数相近的原则确定节流孔径.通过对改造后的EAS试验管线的试验证实,采用本文的设计分析方法设计的多级节流孔板能够有效地减小节流孔板气蚀引起的管道系统振动和噪音.     

The OBE Elimination in Piping Design

The Operating Basis Earthquake (OBE) and Safe Shutdown Earthquake (SSE) have been considered in the design of nuclear power facilities as required by Appendix A to 10 CFR Part 100. However, it is believed that the elimination of the OBE from the design of nuclear facilities would be necessary for plant optimization since the OBE criterion is too rigid and has excessive conservatism. Studies indicate that alternative piping designs can exhibit reliability and safety levels equal to or greater than the current analysis methods. The alternative rules for the Earthquake Engineering Criteria have been issued by the Appendix S to 10 CFR 50. In the System 80⁺ Design, the USNRC reviewed the alternate analysis methods which were proposed to eliminate the OBE based on the EPRI-URD and concluded that those were acceptable as stated in the NUREG-1462. In the Korean Next Generation Reactor (KNGR) developed as an ALWR, a typical piping model was selected to include ASME Classes 1, 2 and 3 piping and was analyzed according to the current method as well as the alternate analysis method, specified in NUREG-1462, for comparison.     

核电厂管系振动鉴定

本文介绍了核电厂管道振动鉴定概况,讨论了目前使用的管系振动的鉴定方法,给出了鉴定准则。     

Piping response testing associated with pipe rupture

EPRI has sponsored an experimental program in the pipe whip impact and pipe rupture and depressurization areas. Sixteen pipe whip tests were performed with 3 in Schedule 80 (or 10) carbon steel pipes impacting on rigid target or concrete slab. The major testing parameters include distance, impact location, pipe rupture location, and concrete slab thickness and strength. The piping crushing at impact correlates with impact force and target response behavior. Conservatism was established by comparing measured and calculated impact forces. The pipe rupture and depressurization tests were carried out using 6 in stainless steel and carbon steel pipes under either PWR or BWR fluid conditions. These tests are of axial crack with initial machined-in surface flaw. It was found that pipe rupture would occur only if a long unstable through-wall crack was embedded in a sufficiently long unstable part-through crack (in the pipe wall). All other flaw configuration tested led to pipe leakage only. Reaction forces were measured which show conservatism of simplified method for fully ruptured condition. No good crack propagation information was obtained.