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.     

A dynamic model for element reliability

A dynamic model is developed for a system element reliability distribution over a generalized strength space. A differential equation is obtained describing the time-dependence of the reliability distribution function (RDF). The equation covers a wide class of power reactor system components which perform under intense stress conditions where a standard subdivision into a “burn-in” period, a “chance failures” range and a “wear-our” period is inapplicable.The hazard distribution function (HDF) over strength is introduced within the model and it is shown that a standard hazard rate is a strength-averaged failure intensity parameter with the RDF as a weighting function.It is shown that a well-known “bathtub” form of the hazard rate function corresponds to an analytical solution of the principal RDF transfer equation under some simplifying assumptions.     

反应堆冷却系统主管道疲劳暨最小壁厚分析方法研究

采用有限元法替代温度场差分方程计算温度瞬态在主管道壁厚方向上的温度分布,将温度计算结果与标准规范的计算公式相结合,从而求解各瞬态交变应力幅,以最终完成先进压水反应堆冷却剂主管道疲劳评定;通过疲劳求解的计算方法研究,提出最小壁厚的优化算法的迭代求解流程,可以依此通过编程最终实现疲劳评价和最小壁厚求解。     

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.     

Development of a PFM code for evaluating reliability of pressure components subject to transient loading

In order to improve the PFM methodology and to evaluate the reliability and integrity of an aged RPV, the PFM code PASCAL is being developed. This code evaluates the conditional probability of crack initiation and fracture of pressure components subject to a transient loading based on Monte Carlo simulation. In addition to the common functions established in existing codes, the code has some original functions and features in the elasto-plastic fracture criterion based R6 method, the simulation models for the semi-elliptical crack extension and the effect of thermal annealing, improvement in Monte Carlo simulation and so on. This paper describes the main features of the code, the results of verification analysis and case studies on influence parameters by using above functions. The verification analysis and case studies are carried based on NRC/EPRI PTS benchmark problem. The basic performance of the code was verified by comparing the results with those by existing codes. From the result of case studies, the effectiveness and performance of main functions are examined and the influence of some parameters, such as fracture criterion, WPS, semi-elliptical crack extension models, existence of overlay clad, initial aspect ratio on the failure probability are also discussed.     

Subassembly experiments and a computer code to analyze the dynamic core deformation during local failure propagation

Safety investigations for LMFBRs have to consider local failure situations in one fuel element which may escalate to a hypothetical CDA. Such initiating events could produce high pressure pulses in a single subassembly which may expand and rupture the wrapper as well as load adjacent elements impulsively. The associated nonlinear dynamic core deformation problem is treated in this paper. In particular the multirow structural dynamics code CØRE-1 and underlying mechanical models are described. Each subassembly is simulated by an equivalent system of point masses and nonlinear coupling springs. The motion of the coolant layer between the elements is treated by an incompressible, non-stationary frictional flow model. In order to obtain realistic code input four types of static single subassembly deformation experiments are described which provided strongly nonlinear load deformation characteristics. Furthermore the transient pressure distribution within the core is obtained from a full scale explosion test. Finally code application is demonstrated and results are given of a transient analysis of the SNR 300 core.     

Quantitative dynamic reliability evaluation of AP1000 passive safety systems by using FMEA and GO-FLOW methodology

The passive safety systems utilized in advanced pressurized water reactor (PWR) design such as AP1000 should be more reliable than that of active safety systems of conventional PWR by less possible opportunities of hardware failures and human errors (less human intervention). The objectives of present study are to evaluate the dynamic reliability of AP1000 plant in order to check the effectiveness of passive safety systems by comparing the reliability-related issues with that of active safety systems in the event of the big accidents. How should the dynamic reliability of passive safety systems properly evaluated? And then what will be the comparison of reliability results of AP1000 passive safety systems with the active safety systems of conventional PWR.

For this purpose, a single loop model of AP1000 passive core cooling system (PXS) and passive containment cooling system (PCCS) are assumed separately for quantitative reliability evaluation. The transient behaviors of these passive safety systems are taken under the large break loss-of-coolant accident in the cold leg. The analysis is made by utilizing the qualitative method failure mode and effect analysis in order to identify the potential failure mode and success-oriented reliability analysis tool called GO-FLOW for quantitative reliability evaluation. The GO-FLOW analysis has been conducted separately for PXS and PCCS systems under the same accident. The analysis results show that reliability of AP1000 passive safety systems (PXS and PCCS) is increased due to redundancies and diversity of passive safety subsystems and components, and four stages automatic depressurization system is the key subsystem for successful actuation of PXS and PCCS system. The reliability results of PCCS system of AP1000 are more reliable than that of the containment spray system of conventional PWR. And also GO-FLOW method can be utilized for reliability evaluation of passive safety systems.     

核级管道的布置和应力分析

核级管道系统的设计是个复杂的优化问题。本文从核反应堆管道系统的设计步骤、管系的解耦准则以及如何通过选择支撑和调整管道走向来降低应力水平等几个方面阐述了此问题。     

GO法在CEFR核级循环冷却水系统的动态可靠性分析

根据CEFR的核级循环冷却水系统的设计和运行特点,基于GO法原理建立了该系统可靠性分析动态计算模型。针对动态分析问题,应用马尔科夫过程理论建立了三种特殊故障模式的部件可靠性参数随时间变化状态转移方程,创建并导出互备自投门的可靠性参数随时间变化的状态转移方程。在此基础上,数值模拟计算了系统的典型运行工况动态可靠性分析。结果表明,该系统的平均可靠度达到0.999 92,说明其设计和设备选型具有很高的可靠性;典型动态分析结果表明,该模型可以合理地计算典型瞬态工况下系统的可靠度随时间的变化。最后对系统进行定性分析。     

Processing dynamic scenarios from a reliability analysis of a nuclear power plant digital instrumentation and control system

This paper illustrates a method for processing accident scenarios generated in a dynamic reliability analysis of a Nuclear Power Plant (NPP) equipped with digital Instrumentation and Control (I&C).The method is based on a Fuzzy C-Means clustering algorithm for classification, which takes into account not only the system states reached at the end of the scenarios but also the timing and magnitude of the occurred failure events, and the characteristics of the process evolution.An illustrative case study is considered, regarding the fault scenarios of the digital I&C of the Lead–Bismuth Eutectic eXperimental Accelerator Driven System (LBE-XADS). A SIMULINK model of the system has been embedded within a Monte Carlo (MC) sampling procedure for injecting faults at random times and of random magnitudes. The accident scenarios thereby generated are classified on the basis of three different system end states, which relate to the value reached by the diathermic oil secondary coolant temperature with respect to maximum and minimum safety threshold values set to avoid primary coolant thermal shocks and degradation of the oil physical and chemical properties.     

Multi-component diffusion analysis and assessment of GAMMA code and improved RELAP5 code

A loss-of-coolant accident (LOCA) has been considered a critical event for very high temperature gas-cooled reactor (VHTR). Following helium depressurization, it is anticipated that unless countermeasures are taken, air will enter the core through the break by molecular diffusion and ultimately by natural convection leading to oxidation of the in-core graphite structure. Thus, without any mitigating features, a LOCA will lead to an air ingress event, which will lead to exothermic chemical reactions of graphite with oxygen, potentially resulting in significant increases of the core temperature.New and safer nuclear reactors (Generation IV) are now in the early planning stages in many countries throughout the world. One of the reactor concepts being seriously considered is the VHTR. To achieve public acceptance, these reactor concepts must show an increased level of inherent safety over current reactor designs (i.e., a system must be designed to eliminate any concerns of large radiological releases outside the site boundary).A computer code developed from this study, gas multi-component mixture analysis (GAMMA) code, was assessed using a two-bulb experiment and in addition the molecular diffusion behavior in the prismatic-core gas-cooled reactor was investigated following the guillotine break of the main pipe between the reactor vessel and the power conversion unit. The RELAP5 code was improved for the VHTR air ingress analysis and was assessed using inverse U-tube and NACOK natural circulation data.     

管道流体瞬态—水汽锤计算原理

文中介绍了核电站管道中流体瞬态——水汽锤的计算原理;既适用于液体介质的水锤计算,也适用于可压汽体的汽锤计算.对于一些典型管道部件的处理方法,文中也作了讨论。     

钍基熔盐堆回路管道系统应力分析与评定

钍基熔盐堆(TMSR)管道设计温度可达700℃,设计标准采用美国机械工程师协会ASME-NH分卷。高温管道评定时除需要进行应力评定外,还需进行应变变形限值和蠕变疲劳限值等评定。利用通用有限元分析软件(ANSYS)对整体回路系统进行计算,并通过优化计算,使得管道应力达到ASME规范中限值要求。     

管道布置验证程序开发及其应用

核电站中的核岛管道布置工作包括管道走向及支吊架设置,需要满足工艺要求、抗震、承重及柔性等要求。主要介绍在台山核电厂项目中一种管道初步设计过程中使用的管道布置验证程序,该程序成功地运用在该设计过程中。结果表明:该程序能够有效地对工艺管道的布置进行力学验证,保证设计质量,减少设计工作的迭代,提高设计效率。     

核电厂商业运行前管道支吊架检查与调整

详细描述在核电厂调试期间对管道支吊架进行冷、热态检查的依据、内容和方法;对检查出有异常支吊架的维修、调整策略及机组临界前热态工况下管道支吊架的验证方式加以说明。提出了支吊架的安装、改造和在役检查中的注意事项,力求通过对管道支吊架的检查与调整,使管道支吊架处于正常的工作状态,保证系统管道安全可靠地运行。     

核电厂反应堆压力容器和主管道焊缝残余应力分析

综述了现有的反应堆压力容器和主管道焊缝残余应力的测试结果和残余应力选取的实践经验。对于反应堆压力容器环焊缝,残余应力沿壁厚呈余弦分布,其最大值可取为60MPa。对于主管道对接环焊缝,最大残余应力区域通常位于在焊缝中心线且靠近管道外表面,而运行过程中的缺陷常出现在内表面区域,在进行安全性评价时焊缝最大残余应力可取为100MPa。