CPR1000蒸汽发生器排污结构设计优化

对中国改进型百万千瓦级压水堆(CPR1000)蒸汽发生器(SG)排污结构进行优化。通过取消排污管及阻挡块,改为在管板上直接开排污孔,提高管廊区域的可达性,便于管板二次侧上表面的检查和泥渣冲洗。应用SG热工水力分析专用软件GENEPI,对比分析优化前后的热工水力特性。结果表明:与原设计方案相比,优化后SG热工水力性能满足设计要求,虽然管板二次侧上表面流场分布发生变化,导致发生泥渣沉积的传热管数量增加,但结构优化后有利于泥渣冲洗,提高冲洗效果。分析结果从理论上证明了优化的可行性。     

蒸汽发生器排污系统的热工-水力特性分析

通过Flowmaster软件建立岭澳核电站3、4号机组的蒸汽发生器排污系统(APG)的水力学模型,改变设置工况,对APG的热工-水力特性进行分析,重点关注对APG影响较大的参数.通过分析可知,由于再生式换热器前的管道具有“热惯性效应”使得蒸汽发生器(SG)温度较快的瞬态波动不会影响到换热器;排污水经再生式热交换器冷却之...     

核电机组大修后主蒸汽压力的变化分析

核电厂的蒸汽发生器(SG)出口蒸汽压力(PSG)也常称为主蒸汽压力,是一个重要的运行参数.在对该参数的监测过程中发现,几次机组大修后PSG都比大修前有所降低,但随着机组的运行,PSG又逐渐恢复到正常的水平.本文针对该现象,首先对这种规律性的变化进行了总结,然后做了一些初步的分析,认为对SG的水力冲洗是造成PSG变化的主要原因,并给出一些改善和提高主蒸汽压力的方法.     

蒸汽发生器柔性泥渣冲洗枪研制

蒸汽发生器(SG)二次侧管板表面堆积的泥渣会随运行时间的增加而变硬,使传热管受到腐蚀和挤压,导致传热管破裂。为了直接对硬泥渣堆积层进行冲洗,研制了一种SG柔性泥渣冲洗枪。本文阐述了SG柔性泥渣冲洗枪的结构、原理、主要技术指标及技术难点。应用结果表明,研制的SG柔性泥渣冲洗枪满足SG柔性冲洗的要求,冲洗效果显著。     

PWR核电站蒸汽发生器停堆湿保养工况联氨的缓蚀作用

一、引言压水堆(PWR)核电站蒸汽发生器(SG)管材因二回路系统腐蚀产物积累发生应力腐蚀开裂,这是SG传热管破损的主要原因之一。维修和更换SG使PWR停运期间所需要的替用电力对发电站造成很大的财政负担,同时,二回路系统的腐蚀产物沉积在蒸汽发生器内,降低     

蒸汽发生器二次侧下部流场与排污试验

位于核电厂蒸汽发生器（SG）管板内的下部排污结构能吸出管板二次侧表面的泥渣并将其排出。为了能合理设计该排污结构并提升排污效率,本文基于非能动大型先进压水堆（CAP1400）的SG设计原型结构,按照1∶4比例设计了排污试验体,以模拟SG下部的管板、传热管等部件。通过对下部流场进行计算流体动力学（CFD）计算并与排污试验的结果进行对比,进一步掌握近管板表面区域的流体流动特征。本试验通过研究SG近管板区域流体流动特征及泥渣分布规律、测量试验体各部件压降、对比SG单边和双边排污结构的设计,为减少淤泥集结、改进设计提供依据。研究发现,单/双边排污结构排污性能基本相同,单边排污结构即可将试验体内泥渣颗粒有效排出。     

压水堆核电厂二回路ETA水化学处理研究

秦山核电厂320 MW核电机组使用乙醇胺(ETA)替换氨作为二回路系统p H调节剂后,在给水p H相同的条件下,汽-水分离再热器(MSR)疏水、蒸汽发生器(SG)排污水的p H明显升高;汽-水两相中水相区域设备的腐蚀产物铁含量明显降低,流动加速腐蚀得到抑制,有效改善二回路系统的腐蚀状况;腐蚀产物向蒸汽发生器二次侧的转移得以降低;同时进一步提高凝水混床的周期制水量,减少了凝水混床树脂的再生次数及再生酸、碱的用量和耗水量,从而减轻运行人员的工作负担和再生废液对环境的污染。     

ACME台架蒸汽发生器传热特性数值模拟

以ACME台架的蒸汽发生器（SG）为研究对象,SG二次侧选用两流体模型,采用计算流体力学软件CFX对ACME台架的SG进行了整体直接模拟。针对稳定试验工况进行了计算,得到了SG一、二次侧的温度分布,二次侧空泡份额分布及传热管的壁温等参数沿U型管高度方向的变化,获得了二次侧较详细的流动和传热特性。计算结果表明,从第2道折流板开始,折流板底部已积聚了部分气泡,随高度的增加,折流板底部积聚的气泡越多,在弯管区附近及以上区域,已全部变为蒸汽。本文计算结果与试验结果符合较好。     

宁德核电N1机组化学性能指标管理实践

宁德核电N1机组二回路水质控制,在不同阶段采取相应的控制措施,系统安装阶段设备保养减少腐蚀、调试和启机阶段严格控制二回路冲洗排污、运行期间逐一排查和消除污染源和有效净化二回路水质等关键过程控制,使得二回路水质得到持续优化,N1机组商运首年化学性能指标达到世界先进水平。     

AP1000蒸汽发生器环焊缝局部热处理防DING分析及监造措施

AP1000蒸汽发生器(SG)的环焊缝局部热处理防永久变形(DING)技术是有别于二代核电SG焊缝局部热处理的一项新工艺,文章从DING产生的原因、影响因素、工艺预防措施进行简单介绍,提出在质量监督过程中需要关注的重要检查点。     

Integrity of the blowdown piping systems in a nuclear power plane-findings and consequences

In a pressurized-water reactor (PWR) the blowdown pipes from the four steam generators (material 1.5415) were examined. All the quality reports of the piping materials and the welding seams as strength and toughness and all non-destructive test findings of the welding seams were listed. According to pipe design drawings, a static calculation of the system under operational loads was conducted. In some cross sections adjacent to elbows an overload was estimated. Selective non-destructive testing of affected sections and inspection of the existing pipe-support conditions were recommended. Non-destructive tests during the inspection in 1994 revealed some longitudinally oriented findings in the base material in the inside walls of two elbows; the two elbows were replaced. The causes of findings were ‘in-plane bending overloading', corrosion processes in the two elbows with findings and, most probably, condensation hammers during the startup period of the plane (possibly the opening of the shutoff valves was too fast) which yielded a large displacement of the piping system. Based on findings, the following should be implemented on the 1996 inspection: (1) the two supports U 27 and U 28 are to be changed; (2) the six highly loaded elbows are to be replaced (wall-thickness is to be increased from 5.6 to 6.3 mm); (3) the whole pipework from the four steam generators to the protection barrier is to be replaced.     

FRAMATOME's continous efforts to improve steam generator corrosion resistance

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.     

Effect of the Quantity, Composition, and Thickness of Deposits on the Damage to the Pipes of Steam Generators in Nuclear Power Plants with VVÉR Reactors

The mechanism of corrosion damage to the heat-transfer pipes of steam generators and the effect of the composition, thickness, and quantity of deposits on the risk of damage to the piping of steam generators are examined.     

Behaviour of the steam generator tubing in water with different pH values

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.     

坎杜堆蒸汽发生器设计综述

国际核电站的运行经验表明，坎杜堆蒸汽发生器具有较高的可靠性，坎杜堆蒸汽发生器的主要型号有：皮克灵Ａ，皮克灵Ｂ，布鲁斯Ａ，布鲁斯Ｂ，根蒂莱２和达灵顿核电站的蒸汽发生器，根据坎杜堆多年的运行经验，其蒸汽发生器设计经历了多次改进，改进后的坎杜型蒸汽发生器使用了能够抵抗各种腐蚀的管材，具有高的循环倍率，柔性支撑板，高效汽水分离装置和方便的维修特性。     

Residual stresses associated with the hydraulic expansion of steam generator tubing into tubesheets

Various methods are being used to expand heat transfer tubes into the thick tubesheets of nuclear steam generators. The residual stresses in the as-expanded tubes and methods for reducing these stresses are important because of the role which residual stresses play in stress corrosion cracking and stress assisted corrosion of the tubing. Of the various expansion processes, the hydraulic expansion process is most amenable to analytical study. This paper presents results on the residual stresses and strains in hydraulically expanded tubes and the tubesheet as computed by two different finite element codes with three different finite element models and by a theoretical incremental analysis method. The calculations include a sensitivity analysis to assess the effects of the expansion variables and the effect of stress relief heat treatments.     

Comprehensive investigation of the corrosion state of the heat exchanger tubes of steam generators. Part I. General corrosion state and morphology

The present work, constituting the first part of a series of two, deals with a systematic investigation of the general corrosion state of 22 heat exchanger tubes originating from different steam generators of the Paks NPP (Hungary). While the passivity of the inner surface of the stainless steel tube specimens was studied by voltammetry, the morphology and chemical composition of the oxide layer formed on the surfaces were analyzed by SEM–EDX method. Based on the measured corrosion characteristics (corrosion rate, thickness and chemical composition of the protective oxide layer) a strong dependence of these parameters on the decontamination history of the steam generators was revealed. It is well documented that the chemical decontamination carried out by a non-regenerative version of the AP-CITROX procedure does exert, on the long run, a detrimental effect on the corrosion resistance of steel surfaces. Therefore, process restrictions and modifications to minimize corrosion damages have be defined.     

蒸汽发生器管子—管板接头设计的计算方法

本文主要介绍液压胀接管子-管板接头设计的一种理论计算方法。该方法是:将接头的整个胀接过程分为四个阶段,以此为数学模形,用弹性理论和塑性理论分阶段对管子或管子和管板的应力、应变和位移进行分析,从而估算胀管压力、变形、残余应力和拔脱力等。该方法可用于设计蒸汽发生器和钢制管壳式换热器。文中指出,对核蒸汽发生器的管子-管板接头来说,采取液压胀后再局部滚胀比只进行液压胀更好。     

核电厂不锈钢覆面焊接模拟件腐蚀行为研究

对3种核电厂乏燃料水池不锈钢覆面材料S32205、S32101和S30403的焊接模拟件,在H₃BO₃浓度2500 mg/L、SO₄2?浓度1500 mg/L、Cl?浓度5%、pH值5.0、温度80℃、饱和氧的条件下浸泡6个月,对比研究其腐蚀行为。结果发现:S30403焊接模拟件在焊接节点和缝隙附近出现了大量的氯致应力腐蚀裂纹;S32101焊接模拟件出现了腐蚀坑,在焊接节点和缝隙附近腐蚀尤其严重;S32205焊接模拟件腐蚀最轻,试件表面未发现腐蚀坑及裂纹。研究表明:3种材料模拟件的耐腐蚀性规律为:S32205>S32101>S30403。S32205具有良好的综合力学性能和耐腐蚀性能,是一种理想的改进型水池覆面材料。      

Experience and status on primary system components after 15 years of operation in the nuclear power plant obrigheim

At Obrigheim the first large pressurized light water reactor built in Germany is operating with a nominal power of 345 MW. Since the beginning of electricity production in later 1968 the nuclear power plant Obrigheim (KWO) has proved a reliable, a safe and also an economical operation with a high availability (83%) over 15 years.KWO has shown that it is possible to prove and maintain the safety and reliability of the primary components on the basis of the present regulations and safety requirements. This was achieved by careful maintenance and by applying improved non-destructive test methodsThe reactor pressure vessel with one circumferential weld near the core could be qualified for future operation by means of inservice inspection, irradiation programs, and by implementation of technical changes for normal as well as for abnormal conditions. To maintain the reliability of the steam generators, extensive eddycurrent testing of the tubes has been performed every year. In order to reduce the corrosive attack on the tubes the secondary water chemistry was controlled very sensitively by minimizing the leakage through the condensor and by using all volatile treatment. The intergranular corrosion of the tubes above the tube sheet could be reduced strongly; but an increasing number of small leakages occurred in the tube sheet region. 458 tubes had been plugged in the old steam generators before they were replaced in 1983 by new ones.In summary it can be stated that the continuous effort to maintain a high quality status of the components is responsible for the high operation availability of the plant.