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蒸汽发生器是核动力装置中一个非常重要的热量交换设备,也是整个核动力装置中的薄弱环节。在核动力装置的停堆事故中,有一半以上是由于蒸汽发生器破损引起的,严重影响到整个核动力装置的安全性和可靠性。如何使传热管不过早破损,延长蒸汽发生器的使用寿命是当前面临的重要课题之一。文章通过对蒸汽发生器传热管所用奥氏体不锈钢材料的分析,结合国内外相关事故分析报告,确定了传热管破损的类型为穿晶型应力腐蚀破坏。通过对腐蚀产物、工作环境以及管束应力的分析,找出蒸汽发生器传热管发生应力腐蚀开裂的影响因素,同时提出相应的防护措施。 相似文献
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蒸汽发生器传热管是反应堆冷却剂压力边界的主要组成部分,这就意味着必须保持传热管的完整性。然而,运行经验表明,蒸汽发生器传热管会出现各种降质。这些降质可能会导致管子的泄漏或破裂,使反应堆冷却剂丧失,并提供了直接通向二回路和释放到环境中去的途径。本文将介绍几种已知的传热管降质,传热管完整性性能准则.并对蒸汽发生器传热管完整性进行评估。 相似文献
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针对蒸汽发生器U形传热管泄漏,本文提出了一种基于时间序列神经网络对蒸汽发生器传热管泄漏程度进行诊断研究的方法。首先,对核电厂蒸汽发生器U型传热管泄漏进行机理分析,构建其数学模型,提取其泄漏的直接特征参数,再依据Fisher得分法,提取其间接特征参数;其次,通过滑动时间窗口法从预处理后的时间序列数据中生成数据样本,作为时间序列神经网络的输入,并以蒸汽发生器U形传热管泄漏程度信息为标注,基于反向传播(BP)算法对五层神经网络系统进行训练,得到蒸汽发生器U形传热管泄漏的时间序列神经网络模型;最后,模拟核电厂运行过程蒸汽发生器U形传热管泄漏时的时间序列测试数据。仿真结果表明,时间序列神经网络对演变事件的处理具有较好的有效性和较高的泛化能力,对故障程度的诊断研究具有参考价值。 相似文献
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快堆管壳式直流蒸汽发生器发生沸腾传热恶化是不可避免的,由此引起的传热管管壁温度波动会使传热管受到疲劳破坏。研究蒸汽发生器的沸腾传热恶化及热疲劳破坏的实验昂贵,难度较大。本文依据国外已发表的实验结果,建立蒸汽发生器沸腾传热恶化发生时传热管管壁温度及热应力的分析模型,应用数值方法求解,对蒸汽压力、质量流速、钠汽温差变动的影响进行了讨论并给出了主要结论。 相似文献
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Dumitra Lucan 《Nuclear Engineering and Design》2011,241(4):1172-1176
Steam generators are crucial components of pressurized water reactors. The failure of the steam generator as a result of tube degradation by corrosion has been a major cause of Pressurized Heavy Water Reactor (PHWR) plant unavailability. Steam generator problems have caused major economic losses in terms of lost electricity production through forced unit outages and, in cases of extreme damage, as additional direct cost for large-scale repair or replacement of steam generators.The excellent performance to date of CANDU steam generators can be attributed, in part, to their design and performance characteristics, which typically involve higher recirculation ratios and lower heat fluxes and temperatures. However, the steam generator tubes are susceptible to failure by a variety of mechanisms, the vast majority of which are related to corrosion.The generalized corrosion is an undesirable process because it is accompanied by deposition of the corrosion products which affect the steam generator performances. It is very important to understand the generalized corrosion mechanism with the purpose of evaluating the quantities of corrosion products which exist in the steam generator after a determined period of operation (IAEA, 1997).The purpose of the experimental research consists in the assessment of corrosion behaviour of the tubes material, Incoloy-800, at normal secondary circuit parameters (temperature—260 °C, pressure—5.1 MPa). The testing environment was the demineralised water without impurities, at different pH values regulated with morpholine and cyclohexylamine (all volatile treatment—AVT).The results are presented like micrographics and graphics representing weight loss of metal due to corrosion, corrosion rate, total corrosion products formed, the adherent corrosion products, released corrosion products, release rate of corrosion products and release rate of the metal. 相似文献
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C. Buchalet 《Nuclear Engineering and Design》1985,86(3)
The steam generators of PWR nuclear reactors are among the primary components most affected by corrosion problems. Corrosion of the steam generator tubes, which assure heat transfer between the primary and secondary circuits, have been observed on a large number of operating steam generators, especially in the United States. According to an NRC survey, as of November 1981, forty PWR units with steam generators of the recirculation type were in operation in the US. Of these, 32 have been found to have one or more forms of tube degradation.Construction of the French PWR nuclear program started in the early 70s, at the time a number of operating plants in the US were being affected by the first corrosion problems. Since, at that time, its construction program was in an early stage, FRAMATOME was able to make modifications on the first units to improve steam generator resistance to corrosion. For instance, full depth expansion of the tubes in the tube-sheet using an explosive process (Westex) was performed on Fessenheim 1 steam generators already installed on site. Later on, continuous operating experience was being obtained in the US, before startup of the French units. This allowed FRAMATOME to react rapidly and take immediate corrective actions at the design stage, during fabrication and sometimes even on site in order to mitigate the risk of corrosion in the steam generators.FRAMTOME is confident that the present design of its steam generator models, including a large number of major improvements is adequate to prevent major corrosion problems to occur during operation. However, the company has embarked on an important development program to further improve the corrosion resistance and thereby the reliability of its steam generators. This program includes studies on new tube expansion techniques, alternate materials for steam generator tubes (Inconel 690), improved tube inspection methods, local thermohydraulic flow, tube vibrations, etc. 相似文献
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Some events of steam generator tubes have been reported in some nuclear power plants around the world. Main causes of the leakage are from various types of corrosion in the steam generator (SG) tubing. Primary water stress corrosion cracking (PWSCC) of steam generator tubing have occurred in many tubes in Korean plants, and they were repaired using sleeves or plugs. In order to develop proper repair criteria, it is necessary to ascertain the leak behavior of the tubings. A high-pressure leak and burst testing system was manufactured. Various types of electro-discharged-machined (EDM) notches having different lengths were machined on the o.d. of test tubes to study SG tube behavior. Leak rate and ligament rupture pressure as well as the burst pressure were measured for the tubes at room temperature. Rupture pressure of the part through-wall defect tubes depends on the defect depth and length. Water flow rates after the rupture were independent of the flaw types; tubes having 20–60 mm long EDM notches showed similar flow rates regardless of the initial defect depth. A fast pressurization rate generated a lower burst pressure than the case of a slow pressurization. 相似文献
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Youn Won Park Myung Ho Song Jin Ho Lee Seong In Moon Young Jin Kim 《Nuclear Engineering and Design》2002,214(1-2)
It is commonly required that steam generator tubes wall-thinned in excess of 40% should be plugged. However, the plugging criterion is known to be too conservative for some locations and types of defects and its application is confined to a single crack. In the previous study, the conservatism of the present plugging criterion of steam generator tubes was reviewed and a crack coalescence model applicable to steam generator tubes with two collinear axial through-wall cracks was proposed. Since parallel axial cracks are more frequently detected during in-service inspections than collinear axial cracks, the studies on parallel axial cracks spaced in circumferential direction are necessary. The objective of this paper is to investigate the interaction effect between two parallel axial through-wall cracks existing in a steam generator tube. Finite element analyses were performed and a new failure model of the steam generator tube with these types of cracks is suggested. Interaction effects between two adjacent cracks were investigated to explain the deformation behavior of cracked tubes. 相似文献
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PWR核电站蒸汽发生器传热管和主管道的应力腐蚀破裂研究 总被引:2,自引:0,他引:2
用慢应变速率试验(SSRT)、恒载荷试验(CLT)和低周循环载荷试验方法研究以秦山和大亚湾核电站安全为目的的有关压力边界管道破裂始发事件应力腐蚀破裂(SCC)的行为,为评价管道的结构完整性提供支持性实验数据。研究的材料有核等级主管道焊接热影响区(WHAZ)316不锈钢(SS),核等级蒸汽发生器(SG)传热管材Incoloy-800、Inconel-600、Inconel-690和321SS。研究的影响因素包括材料冶金、表面喷丸处理、载荷、应变速率、循环载荷以及水化学条件对SCC的影响规律。 相似文献
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Hyun-Su Kim Tae-Eun Jin Hong-Deok Kim Yoon-Suk Chang Young-Jin Kim 《Nuclear Engineering and Design》2008,238(1):135-142
The steam generator in a nuclear power plant is a large heat exchanger that uses heat from a reactor to generate steam to drive the turbine generator. Rupture of a steam generator tube can result in release of fission products to environment outside. Therefore, an accurate integrity assessment of the steam generator tubes with cracks is of great importance for maintaining the safety of a nuclear power plant. The steam generator tubes are supported at regular intervals by tube support plates and rotations of the tubes are restrained. Although it has been reported that the limit load for a circumferential crack is significantly affected by boundary condition of the tube, existing limit load solutions do not consider the restraining effect of tube support plate correctly. In addition, there are no limit load solutions for circumferential cracks in U-bend region with the effect of tube support plate. This paper provides detailed limit load solutions for circumferential cracks in top of tube sheet and the U-bend regions of the steam generator tube with the actual boundary conditions to simulate the restraining effect of the tube support plate. Such solutions are developed based on three-dimensional (3-D) finite element analyses. The resulting limit load solutions are given in a polynomial form, and thus can be simply used in practical integrity assessment of the steam generator tubes. 相似文献
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