核电厂混凝土安全壳中预应力的分布和损失

本文研究了核电厂安全壳预应力系统建立过程中混凝土的应力值、安全壳应力分布模式和由于预应力施加产生的变形情况，并把这些数据与在安全壳结构强度试验（SIT）中得到的值进行比较分析，通过理论计算，讨论安全壳中预应力损失以及其安全性问题。     

核电厂安全壳中预应力的数值模拟

在预应力混凝土安全壳结构计算中,预应力的计算分析以及模拟是十分重要的一部分。本文根据某核电厂安全壳预应力的布置情况,对预应力损失的分析过程进行了说明,并介绍了在安全壳数值模拟中用降温法模拟预应力的具体方法,同时采用修正系数对温降值进行修正,消除了传统一次降温法所产生的预应力损失,使预应力的模拟更为精确。此方法具有较高的通用性,供行业内工程设计人员参考。     

考虑预应力损失影响的安全壳安全性能研究

预应力损失对安全壳在内压作用下的安全性能影响不可忽略，本文通过考虑安全壳不同龄期下的预应力损失来研究安全壳在设计基准期内40年及设计基准期后60年不同内压水平作用下的安全性能。采用ABAQUS有限元软件建立了精细化安全壳三维有限元分析模型，通过非线性有限元方法分析了钢衬里屈服、预应力筋屈服、混凝土裂缝演化等性能指标。研究结果表明，考虑预应力损失后，安全壳混凝土开裂与钢衬里失效时，所能承受的内压荷载减小；安全壳在极限内压作用下的变形表现为穹顶向外膨胀以及洞口向内收缩；安全壳穹顶部分在极限内压下破坏严重；考虑预应力损失后，安全壳变形明显增大。但安全壳在设计内压(0.4 MPa)作用下仍有足够的安全裕度。     

混凝土安全壳整体性能试验峰值压力持续时间探讨

核电站建造阶段必须进行安全壳整体性能试验（CTT）,验证在设计基准事故时安全壳结构的完整性。本文针对某核电厂3号机组预应力混凝土安全壳CTT进行非线性有限元分析。结果表明：筒体闸门洞口标高附近径向变形最大,预应力钢束承担了峰值压力0.483 MPa作用下大部分设计内压,安全壳整体结构处于受压状态,与实际试验状态基本吻合。同时,对国内外法规标准关于安全壳峰值压力持续时间的规定进行总结,提出相关结论及建议,可为安全壳CTT方案设计提供参考。     

基于构件尺度的预应力混凝土徐变模型研究

为评估核电厂安全壳结构的长期预应力损失,以预应力混凝土梁为研究对象,采用试验研究与理论分析相结合的方法,建立预应力混凝土徐变预测模型。在已有的预应力混凝土梁徐变试验基础上,采用相同的混凝土材料进行相同环境下的收缩试验,以测定预应力混凝土梁的实际收缩变形。考虑到混凝土收缩、徐变、预应力筋松弛的耦合作用,引入龄期调整有效模量法,建立由试验数据推导混凝土徐变系数的计算方法,最终建立预应力混凝土徐变模型并预测其长期徐变变形,为核电厂安全壳结构长期预应力损失评估提供了理论支撑。     

基于ANSYS的核电厂安全壳结构非线性有限元分析

对核电厂预应力混凝土安全壳结构进行了内压作用下的非线性有限元分析.详细介绍了ANSYS中的混凝土单元SOLID65及混凝土材料的本构关系,并对非线性求解过程中影响收敛的因素进行了分析；同时,以福清核电厂5、6号机组内层安全壳为工程实例进行有限元计算.结果表明,15 m至30 m标高范围内的径向位移大于其他高度的径向位移,标高25 m左右径向位移最大；内压加至0.42MPa,模型结构仍处于受压状态,满足使用要求.分析表明,福清核电厂5、6号机组安全壳结构在设计内压作用下是安全的,可为安全壳整体性试验提供参考.     

某核电厂LOCA下预应力混凝土安全壳响应规律初探

核电厂LOCA发生后,预应力混凝土安全壳结构内温度场分布具有明显的非线性特征,但现行的混凝土安全壳设计规范未对LOCA下温度和应力的组合作用提出具体的计算方法。基于用ANSYS程序建立的包含预应力钢束的混凝土安全壳结构的有限元模型,本文计算了LOCA下不同时刻安全壳壳壁内的温度场分布,并与理论值进行了比较,验证了计算模型的正确性。初步分析了高温、高压作用下安全壳结构变形的规律,总结了混凝土温度效应和预应力系统的作用,可为安全壳结构设计提供参考。     

百万千瓦级核电厂安全壳结构设计与试验研究

通过建立符合先进核电厂安全壳结构特点的线性和非线性有限元分析模型,得出合理的安全壳预应力张拉顺序,计算出安全壳在设计事故内压、严重事故内压状态下的工作生能及其极限承载能力,并与110的大比例尺结构模型试验结果相互比较,取得一致的结论先进核电厂安全壳符合国际上极限承载力≥25倍设计内压的合格标准.从而验证了先进核电厂安全壳概念设计的合理性.     

先进核电厂半球顶安全壳抗震分析

安全壳是核电厂反应堆主厂房的围护结构,是防止设计事故发生时放射性物质扩散的最后一道屏障,是确保核电厂安全的关键设施.因此,必须在设计中考虑到安全壳在可能的、会引发重大核事故的意外荷载作用下的工作性能.地震是核电厂整个使用过程中有可能出现的自然灾害之一,并可能引发重大事故,所以,必须对安全壳结构进行严格的抗震性能分析,设计要保证预应力混凝土安全壳能够承受SSE作用而不被损坏.本文通过有限元模型的计算与分析,得到先进核电厂半球顶安全壳结构在SSE作用下的应力、变形、位移等地震反应,由此进行安全壳结构构件抗震分析计算.计算表明,半球顶安全壳结构在SSE作用下,安全壳结构安全可靠,结构的设计能够满足我国核电厂安全导则对抗震Ⅰ类结构的规定.     

先进核电厂安全壳结构模型试验

安全壳是承受设计基准事故工况的安全屏障。掌握其先进技术，对于保证我国的核电地位，适应核电技术发展方向有着重要意义。冶金工业部建筑研究总院受上海核工程研究设计院的委托，承担了先进核电厂安全完结构模型试验项目。安全壳结构模型试验与工程实体安全壳结构试验相比，可对关键技术问题进行更加深入的试验研究。本模型试验通过位移、应变、裂缝、预应力值的试验结果，检验结构的非线性分析理论，实测极限承载能力和破坏状态，并为实体设计提供试验依据。最终的破坏试验将在今年7月结束。本模型属于第三代预应力混凝土安全壳，模型与…     

Reliability-based assessment of integrity of bonded prestressed concrete containment structures

The main function of a nuclear containment structure is to prevent the leakage of radioactive materials from the reactor in the event of a serious failure in the process system. To maintain a high level of leak integrity, prestressed concrete is widely utilized in containment construction. In bonded prestressing systems, excessive prestressing losses caused by unexpected material deformations and degradation of tendons could result in the loss of leak integrity under an accident. To safeguard against this, the Canadian Standard, CSA N287.7 (1995), recommends periodic inspection and evaluation of prestressing systems of CANDU containments. As bonded tendons are not amenable to direct inspection, the evaluation is based on the testing of a set of beams with features identical to the containment. The paper presents a quantitative reliability-based approach to evaluate the containment integrity in terms of the condition of bonded prestressing systems. The proposed approach utilizes the results of lift-off, destructive, and flexural tests to update the probability distribution of prestressing force, and to revise the calculated reliability against through-wall cracking of containment elements. An acceptable criterion for the results of beam tests is established on the basis of maintaining adequate reliability throughout the service life of the containment.     

Containment passive-cooling design concepts

A containment is proposed for a high rating PWR (1300 MW_e) that makes it possible to reject sufficient heat to maintain internal conditions below design limits during any postulated design basis accident. The proposed containment thus eliminates the need to employ active features for containment cooling, and conformes to guidelines set forth for passive reactor systems. [EPRI, 1987] The design is based in part on a currently operating PWR containment (Waterford 3). A series of modifications and additions are necessary to make passive heat rejection possible. The modifications are an increase in free volume and primary shell surface area. The additions are the perforation of the secondary containment structure to form an air-convection annulus, and allow the submersion of the lower part of the containment into an external pool; an internal pool increases in-containment heat storage. Proposed features are evaluated analytically, computationally and, where possible and necessary, experimentally. The proposed containment is shown to remain below current regulatory limits for the design-basis postulated loss of coolant accident.     

Thirty years of measured prestress at Swedish nuclear reactor containments

The main function of the reactor containment, i.e. to ensure tightness at a major internal accident, depends directly on the prestressing system. To secure that the prestress level is sufficient, the tendon force has been measured during the whole time of operation. The general results from these measurements show that the loss of prestress 30 years after tensioning is between 5 and 10%. This is much lower loss than predicted initially at the design stage. More advanced and today commonly used models for predicting prestress loss show better agreement with the results. The main reasons for the relatively low loss are assumed to be: (1) the confirmed slow drying process of the concrete and (2) the high concrete age at the initial tensioning. The results also indicate that the temperature has a major influence on the loss of prestress.     

Conceptual design and analysis of a semi-passive containment cooling system for a large concrete containment

An internal evaporator-only (IEO) concept has been developed as a semi-passive containment cooling system for a large dry concrete containment. The function of this system is to keep the containment integrity by maintaining the internal pressure not to exceed ultimate design pressure, i.e. 0.83 MPa (120 psia) in the absence of any other containment cooling following a severe accident, which postulates core damage and hydrogen combustion. The ability of the concept to protect the containment was evaluated for the design basis accident (DBA) large break loss of coolant accident (LB LOCA) and severe accident scenarios (LB LOCA without Emergency Core Cooling System (ECCS) and containment spray flow, 100% zirconium oxidation and complete hydrogen combustion). All were modeled using the GOTHIC computer code. It was concluded that a practical system requiring four IEO loops could be utilized to meet design criteria for severe accident scenarios.     

Prediction of radial stresses due to prestressing in PSC shells

Large prestressed concrete (PSC) shells, such as silos, containment structures are prestressed along their principal curvatures. Prestressing cables are provided in the thickness of the shell either in one or two directions. The stresses normal to the middle surface are generally neglected in the design of such shell structures. However, the development of radial tensile stresses at the time of prestressing can lead to possible delamination failure, if these stresses are excessive. This paper presents simple procedures to estimate the radial stress distribution due to prestressing. Two methods are presented: ‘equilibrium of slice’ and ‘modified Lamé,’ the latter being more rigorous and applicable to both singly and doubly curved shells. The solutions obtained are in conformity with the results of finite element analysis published earlier. Finally, an attempt is made to determine the local tensile stress distribution at the periphery of the prestressing duct hole, which also contributed to the problem of delamination.     

Creep prediction of concrete for reactor containment structures

Since the biggest time-dependent prestress loss of a prestressed concrete nuclear reactor containment structure is due to the creep of concrete, creep is one of the most important structural factors to be considered for the safety of a reactor containment structure during design, construction and maintenance. Creep in concrete has also recently been considered in evaluation of the crack resistance of concrete at an early-age in the durability examination of massive concrete structures like reactor containment structures. Existing empirical formulas on creep prediction show errors in their predictions due to simplified consideration of mixture proportions, and they also show large discrepancy among their predictions. In addition, they do not consider early-age behaviors of concrete and thus are mainly for the prediction of long-term creep at hardened concrete. In this paper, the creep characteristics of the reactor's both early-age and hardened reactor concrete made of type V cement are examined by carrying out both early-age and long-term creep tests. Then, the creep of the reactor concrete is predicted by using major creep-prediction equations of the AASHTO LRFD design specification, the Japanese standard specification for concrete structure, the ACI Committee 209 and the CEB/FIP model code and the Bazant and Panula's model, and the predicted results are compared with the test results. From the comparison, the applicability of the creep-prediction equations for the concrete of a reactor containment structure at both early-age and hardened stages is discussed.     

严重事故条件下安全壳响应模拟研究

压水堆核电厂发生严重事故期间，从主系统释放的蒸汽、氢气以及下封头失效后进入安全壳的堆芯熔融物均对安全壳的完整性构成威胁。以国内典型二代加压水堆为研究对象，采用MAAP程序进行安全壳响应分析。选取了两种典型的严重事故序列：热管段中破口叠加设备冷却水失效和再循环高压安注失效，堆芯因冷却不足升温熔化导致压力容器失效，熔融物与混凝土发生反应（MCCI），安全壳超压失效；冷管段大破口叠加再循环失效，安全壳内蒸汽不断聚集，发生超压失效。通过对两种事故工况的分析，证实了再循环高压安注、安全壳喷淋这两种缓解措施对保证安全壳完整性的重要作用。