基于荷载时程分析法的商用飞机撞击钢板混凝土结构安全壳的有限元分析

由于钢板混凝土墙背部钢板能够有效地约束混凝土在撞击方向上的运动以及限制混凝土碎片的飞溅,为了抵御商用飞机撞击,新型核电机组的核岛厂房外墙通常设计为钢板混凝土结构(SC)。本文基于荷载时程分析法,运用经典的显式非线性动力分析软件ANSYS/LS-DYNA,进行了波音707-320型号商用飞机撞击某钢板混凝土结构安全壳的响应分析。计算结果表明,即使在安全壳筒身最不利撞击部位冲击作用下,像波音707-320型号的商用飞机对该安全壳的影响是较小的,且增大钢板的厚度能够有效的减小冲击作用下结构的响应。     

具有隔震构造的核电站安全壳在强烈地震和大型商用飞机撞击下的振动响应分析

核电站安全壳的安全性和完整性在建造运营期间必须要得到保证,尤其是对于强烈地震和飞机撞击等超设计基准事故。基础隔震作为一种能有效降低上部结构地震响应并防止结构破坏的典型隔震措施,在未来的核电站设计建造中备受关注,因而在核电站隔震系统的建造设计时应考虑大型商用飞机撞击的影响。该文基于CPR1000核电站安全壳,在底板下侧布置了不同型号的高阻尼橡胶隔震支座,从结构的隔震率（DAR）、最大水平位移和加速度反应谱,对安全壳在强烈地震作用下的隔振性能进行了综合分析。并结合采用荷载时程法以A320商用飞机撞击荷载为例进行飞机撞击核电站安全壳的分析,着重从飞机撞击后的振动衰减过程的振动特性以及飞机撞击诱导振动的加速度反应谱的分析结果对具有隔震构造的安全壳振动规律进行了分析。     

AP1000安全壳在大型商用飞机撞击下的动力响应分析

自“9·11”事件以来,核电站抵御大型商用飞机撞击一直是核安全领域的热点问题。采用ANSYS/LS-DYNA软件建立波音737 MAX 8和AP1000安全壳的精细化有限元模型,基于Riera法验证了飞机撞击有限元模型准确性,进行了5种不同初始撞击速度(100 m/s, 150 m/s, 200 m/s, 250 m/s和300 m/s)和5种不同撞击高度(30 m, 39 m, 47 m, 54 m和65 m)的飞机撞击安全壳全过程数值模拟,研究和分析了飞机的撞击力时程和动能时程、钢制安全壳的动力响应、等效应力分布和其局部破坏情况。结果表明：飞机撞击过程引擎对撞击力的贡献最大,约为机身撞击力的3～4倍;筒身等效钢梁处的撞击力峰值较非等效钢梁处大,最大达到了后者的171%(速度为300 m/s);安全壳筒身段与穹顶的交界处为安全壳结构的最危险位置,在此位置处安全壳环向和竖向贯穿尺寸均大于同初始撞击速度下其他位置的贯穿尺寸,最大环向、竖向贯穿尺寸分别达到29.68 m和17.86 m;当飞机的撞击速度大于150 m/s时,撞击区域钢板等效应力影响范围随初始撞击速度的增加而减少,撞击等效钢...     

大型商用飞机撞击刚性墙及核电屏蔽厂房的撞击力分析

建立一个综合考虑了真实质量分布、刚度分布和材料模型的Boeing767-200ER模型,利用LS-DYNA程序对其撞击刚性墙进行数值模拟,获得了飞机撞击荷载时程曲线,并分析了不同初始速度及内部结构对飞机撞击力的影响;将该飞机有限元模型用于撞击核电站屏蔽厂房的数值模拟中,分析了靶体形状及相对刚度对飞机撞击力的影响,为飞机撞击核电站问题的深入研究奠定了基础。     

飞机撞击荷载计算模型中压溃力选取分析

Riera提出的飞机撞击荷载计算模型是一种比较实用的简化模型,大多数研究在进行飞机撞击试验和数值模拟时通常会把试验结果和数值模拟结果与该模型的计算结果进行对比分析。通过介绍分析Riera模型,该模型中撞击荷载包括压溃力和惯性力两部分,因此模型的使用需要确定压溃力大小。为了合理使用Riera模型,梳理了现有压溃力计算公式,各公式给出的压溃力计算结果差别较大,并通过公式推导预测弹体的变形长度和撞击作用时间与试验数据进行对比分析,评价了现有压溃力计算公式,对更适宜于飞机撞击条件下的压溃力计算公式给出了说明,最后讨论了压溃力对撞击荷载曲线的影响,得出压溃力相对于惯性力不能被忽略,对撞击荷载影响较大的结论。     

AP1000安全壳在大型商用飞机撞击下的动力响应分析EI北大核心CSCD

自“9·11”事件以来,核电站抵御大型商用飞机撞击一直是核安全领域的热点问题。采用ANSYS/LS-DYNA软件建立波音737 MAX 8和AP1000安全壳的精细化有限元模型,基于Riera法验证了飞机撞击有限元模型准确性,进行了5种不同初始撞击速度(100 m/s,150 m/s,200 m/s,250 m/s和300 m/s)和5种不同撞击高度(30 m,39 m,47 m,54 m和65 m)的飞机撞击安全壳全过程数值模拟,研究和分析了飞机的撞击力时程和动能时程、钢制安全壳的动力响应、等效应力分布和其局部破坏情况。结果表明:飞机撞击过程引擎对撞击力的贡献最大,约为机身撞击力的3~4倍;筒身等效钢梁处的撞击力峰值较非等效钢梁处大,最大达到了后者的171%(速度为300 m/s);安全壳筒身段与穹顶的交界处为安全壳结构的最危险位置,在此位置处安全壳环向和竖向贯穿尺寸均大于同初始撞击速度下其他位置的贯穿尺寸,最大环向、竖向贯穿尺寸分别达到29.68 m和17.86 m;当飞机的撞击速度大于150 m/s时,撞击区域钢板等效应力影响范围随初始撞击速度的增加而减少,撞击等效钢梁处的钢板等效应力分布范围相对于撞击非等效钢梁处的更大。研究成果可为类似核岛安全壳的抗飞机冲击安全评估和设计提供参考。     

撞击荷载下钢筋混凝土柱动力响应的数值研究

为了研究撞击荷载下钢筋混凝土柱的动力响应,该文基于有限元软件LS-DYNA建立了钢筋混凝土构件碰撞模拟分析的数值模型。对经典钢筋混凝土梁落锤试验进行数值模拟,分析表明该数值模型能够较好的模拟钢筋混凝土构件碰撞过程中的撞击力和变形发展。基于该模型该文研究了不同因素对刚体碰撞钢筋混凝土柱动力变形和撞击力的影响,包括撞击体速度、质量和形状,钢筋混凝土柱纵向配筋率、配箍率和混凝土轴心抗压强度以及材料应变率。分析表明钢筋混凝土柱在撞击作用下的破坏模式主要包括：局部损伤型、整体破坏型和局部破坏型。影响撞击力曲线变化的因素可分为三种：撞击体自身的因素,影响柱体刚度的因素和影响柱体材料强度的因素。各种因素对撞击力峰值、撞击力平台值和撞击持时分别具有不同的影响规律。这些规律将为建立建筑结构倒塌分析的简化撞击力模型提供参考。     

飞机撞击特大型LNG储罐全过程仿真分析

采用数值模拟方法对飞机撞击特大型LNG储罐的全过程进行仿真分析。分析中采用LS-DYNA有限元程序,考虑罐体、储液与保温层间的相互问题,建立了F-15战斗机的SPH模型,对飞机材料的选择和参数确定进行了详细分析,并以Riera法为依据,对F-15战斗机SPH模型撞击刚体所产生的荷载进行了对比验证,对比结果证明了SPH模型的可靠性和实用性。分析结果表明：撞击角度越大,外罐所承受的撞击能量越大,相应的内罐破坏越小,因此垂直撞击为最不利撞击角度;撞击高度对整体工况计算结果影响不大,储罐在经受215m/s撞击速度撞击下均出现了严重破坏;112m/s撞击速度时内罐尚有安全余量,160m/s撞击速度时内罐撞击中心区域内材料已达到极限应变,因此可认为目前设计方法设计的储罐所能承受的最大撞击速度为160m/s。     

飞机撞击下核岛厂房损伤破坏与振动响应的数值仿真分析EI北大核心CSCD

9·11事件后,商用飞机撞击成为核岛厂房必须考虑的超设计基准事故。为探讨合理可靠的数值仿真方法以准确评估核岛厂房在商用飞机撞击下的损伤破坏和振动响应,首先对1∶15缩尺飞机撞击核岛厂房模型试验进行数值模拟。通过对比试验中厂房结构的损伤破坏以及不同标高位置的加速度时程,验证了所采用的材料模型与参数,以及数值仿真方法的准确性。进一步建立了精细化的AP1000原型核岛厂房有限元模型,开展了A320飞机以100 m/s的速度撞击反应堆厂房环吊梁标高位置的数值模拟,评估了厂房的损伤破坏和环吊梁的振动响应。结果表明:在该研究的工况中,商用飞机撞击反应堆厂房将导致机身撞击区域发生贯穿破坏,但内部钢制安全壳可保持整体密封性;环吊梁振动低于限值,环吊不会发生跌落,反应堆安全性可以保证。     

混凝土浇筑期泵管撞击荷载的试验研究

由泵管往复运动撞击布料杆而产生的泵管撞击荷载是作用在模板支架上水平动荷载的诱发荷载,当前对这一荷载还知之甚少。该文对18个施工现场进行调研,基于调研结果,在3个工地对泵管的往复运动进行了测试,掌握了泵管运动的特点;根据达朗伯原理,给出了泵管撞击荷载的估算公式;对布料杆加速度和布料杆相对位移进行了测试,给出了泵管撞击力的变化区间。该文研究成果为研究作用在模板支架顶部的水平动荷载打下基础。     

A study of a commercial MHD power plant scheme

Power engineering specialists are currently interested in electrical power stations with magnetohydrodynamic generators. This interest has been generated by the fact that fossil fuels are becoming increasingly costlier, and with the exploitation of remote and practically inaccessible deposits, a more rational utilisation of fuel has become necessary. Research on magnetohydrodynamic generators is being conducted in many countries at present. In the USSR a composite pilot plant with an MHD generator whose output exceeds 20 MW has already been operative for several thousand hours. However the pilot plant has to be considerably modified to serve as a model for the 500–1000 MW industrial power unit. This paper is devoted to an investigation of one of the possible process flow diagrams of MHD electrical power plants. The structure of MHD electrical power plants, the interrelation between the aggregates, issues concerning the starting of the plant and the working of the power unit under various partial load conditions are discussed. With the availability of new theoretical and experimental data, the process flow diagrams of industrial MHD electrical power plants will naturally undergo changes. However, the methodical approach and the investigations, described in this paper should retain their validity for all process flow diagrams of electrical power plants with MHD generators.     

An analysis of maintenance failures at a nuclear power plant

In this paper, a study of faults caused by maintenance activities is presented. The objective of the study was to draw conclusions on the unplanned effects of maintenance on nuclear power plant (NPP) safety and system availability. More than 4400 maintenance history reports from the years 1992–1994 of Olkiluoto BWR NPP were analysed together with the maintenance personnel. The human action induced faults were classified, e.g. according to their multiplicity and effects. This paper presents and discusses the results of a statistical analysis of the data. Instrumentation and electrical components appeared to be especially prone to human failures. Many human failures were found in safety related systems. Several failures also remained latent from outages to power operation. However, the safety significance of failures was generally small. Modifications were an important source of multiple human failures. Plant maintenance data is a good source of human reliability data and it should be used more in the future.     

核电站检修主从式水下作业机械手研究

研制了适用于核电站反应堆和乏燃料水池水下作业的六自由度同构型主从式水下机械手。建立了从机械手水动力学模型,针对主从式机械手具有的强非线性,设计了一种基于模糊PID控制的主从控制器。采用ds PACE半物理仿真平台,搭建了主从式机械手两关节测试系统,验证了主从实时跟随性,为水下作业机械手整机控制奠定基础。实验结果表明主从手关节角度跟随性好,主从手末端位置跟随过程形成的轨迹形状基本一致,大小成比例,证明模糊PID控制器能较好地实现主从轨迹跟随,满足位置伺服精度要求。     

A framework for the selection of the right nuclear power plant

Civil nuclear reactors are used for the production of electrical energy. In the nuclear industry vendors propose several nuclear reactor designs with a size from 35–45?MWe up to 1600–1700?MWe. The choice of the right design is a multidimensional problem since a utility has to include not only financial factors as levelised cost of electricity (LCOE) and internal rate of return (IRR), but also the so called “external factors” like the required spinning reserve, the impact on local industry and the social acceptability. Therefore it is necessary to balance advantages and disadvantages of each design during the entire life cycle of the plant, usually 40–60 years. In the scientific literature there are several techniques for solving this multidimensional problem. Unfortunately it does not seem possible to apply these methodologies as they are, since the problem is too complex and it is difficult to provide consistent and trustworthy expert judgments. This paper fills the gap, proposing a two-step framework to choosing the best nuclear reactor at the pre-feasibility study phase. The paper shows in detail how to use the methodology, comparing the choice of a small-medium reactor (SMR) with a large reactor (LR), characterised, according to the International Atomic Energy Agency (2006), by an electrical output respectively lower and higher than 700?MWe.     

Behavioural analysis of a hydroelectric production power plant under reworking scheme

By high calibre of the fundamental nature of electric power, both of our economic and personal interests, a system is expected to supply as electrical energy as possible with the highest degree of quality and reliability. This research work aimed to evaluate the reliability measures of a hydroelectric power station. The results of this study are deliberate to provide the improved criteria for upcoming proposals, and serves as a basis for generation expansion planning of hydroelectric power stations. This paper presents a Markov-process based mathematical model for a hydroelectric power plant. The important reliability measures have been derived and discussed with the help of failure and repair rates through the analysis. This research gives a clear view of how the hydropower plants are modelled by defining the states under the failure and repair rates.     

核电厂系统人因工程学的应用研究

文章阐述了核电厂人因工程学应用现状，并介绍了应用在核电厂的人的可靠性分析的方法，提出了应用人因工程学改进核电厂可靠性与安全性的一些建议。     

Analysis of operators' performance under emergencies using a training simulator of the nuclear power plant

It is well known that there are many factors that affect the reliability of nuclear power plants (NPPs). Among them, human reliability has been considered one of the most important factors. Thus, not only in order to quantify human reliability but also to identify main causes that can degrade human reliability, various kinds of human reliability analysis (HRA) methods have been suggested and utilized in many countries. However, to perform HRA more appropriately, it is necessary to collect plant-specific or domain-specific human performance data: especially for emergencies: because they can be used to generate requisite information for HRA. From this point of view, simulator studies under emergencies may be considered important sources for obtaining human performance data.In this study, the performance data of operating crews in coping with emergencies of the reference NPP have been collected and analyzed to develop human performance database (HPDB). Since the number of collected records is 112, it can be said that extracted/analyzed results included in HPDB are statistically meaningful. Therefore, HPDB can be used not only for HRA input data but also for multiple purposes such as improving emergency operating procedures and developing advanced HRA methods.     

Performance test and analysis after high pressure turbine retrofit for Maanshan nuclear power plant Unit 1

The purpose of the performance test for Unit 1 of Maanshan nuclear power plant was to determine the electrical output and heat rate after the retrofit of the high-pressure turbine during the refueling in 2012. The performance test was conducted in order to verify that the actual improvement in electrical output resulting from the replacement of the high-pressure turbine meets the vendor’s guarantee. A total of two performance test runs was conducted in accordance with the American Society of Mechanical Engineers performance test code (PTC) 6. The measured electrical powers for the two test runs were 977.4 and 975.0 MWe, respectively, and the average value was 976.2 MWe. After correcting the electrical power to the rated conditions specified in the performance test procedure, the gross electric output was 983.2 MWe. The corrected heat rate for the two performance tests were 10365 and 10353 kJ/kWh, respectively. The deviation between two corrected heat rates was 0.11%. Since the acceptable deviation between two test runs required by PTC 6 is no more than 0.25%, the quality of test results is acceptable. Moreover, the performance test results also demonstrated that the improvement in gross electrical output was 17.6 MWe, which was higher than the contract guarantee of 10.0 MWe.     

Feasibility analysis of an OTEC plant as the bottom cycle of the third nuclear power plant

Abstract

A conceptual design, performance analysis, system optimization and simulation of an OTEC (Ocean Thermal Energy Conversion) power plant was carried out in the present study. The OTEC plant acts as the bottom cycle of the Third Nuclear Power Plant, utilizing the warm water discharged from the nuclear plant as the heat source and the cold deep seawater pumped from the nearby sea bottom as the heat sink. The nominal net power output from the OTEC plant is estimated to be 8.74 MWe. A plant simulation was further carried out to study the OTEC plant performance due to seasonal variation of the warm water temperature, which shows that the OTEC plant is able to deliver an electricity of 5.18×10⁷ kWH per year. The net power output in the winter season will be reduced to about one eighth of that in the summer season. The final discharged warm water temperature was shown to be below 29°C all year around, so that the current thermal pollution problem of the nuclear power plant can be eliminated. Additionally there is the side benefit of 8.74 MWe net power generation from ocean energy resource. This OTEC plant will become economically feasible if the installation cost is considered as a thermal pollution control investment.     

Assessment of human–machine interface design for a Chinese nuclear power plant

A nuclear power plant (NPP) is a complex system but requires high reliability. The human–machine interface (HMI) design plays very important role in reactor safety. This paper describes an assessment on HMI design of a Chinese NPP, using a software system named Dynamic Interaction Analysis Support (DIAS). DIAS can give not only quantitative indices for dynamically assessing the HMI design, but also allow modify the values of these indices by taking into account human error probability during specified emergent operation procedures. The operation procedures dealing with postulated accidents and transients recorded from a full-scale plant simulator in the training center of a Chinese NPP were selected as references. According to the results of simulation and analysis, the potential problems in the HMI design and the operation procedures were detected. Suggestions to improve the HMI design and the operation procedures were addressed.