Review on synergistic damage effect of irradiation and corrosion on reactor structural alloys

The synergistic damage effect of irradiation and corrosion of reactor structural materials has been a prominent research focus.This paper provides a comprehensive review of the synergistic effects on the third-and fourth-generation fission nuclear energy structural materials used in pressurized water reactors and molten salt reactors.The competitive mechanisms of multiple influencing factors,such as the irradiation dose,corrosion type,and environmental temperature,are summarized in this paper.Conceptual approaches are proposed to alleviate the synergistic damage caused by irradiation and corrosion,thereby promoting in-depth research in the future and solving this key challenge for the structural materials used in reactors.

Corrosion of candidate materials for supercritical water-cooled reactor

Supercritical water reactor(SCWR) was proposed as a GenerationⅣconcept for building large capacity nuclear power plants.Comparing with the present GenerationⅡandⅢlight water reactors,SCWR possesses great advantages of 10%higher efficiency,simpler system design,better sustainability,and so on. However,the selection of materials for fuel cladding and reactor internals of SCWR is facing a great challenge. Corrosion in supercritical steam is of the first important issue to be solved to meet the stringent requirement of the reactor internal components.Corrosion screening tests were conducted on candidate materials for nuclear fuel cladding and reactor internals of supercritical water reactor(SCWR) in static and re-circulating autoclave at the temperatures of 550,600 and 650℃,pressure of about 25 MPa,deaerated or saturated dissolved hydrogen(STP). Nickel base alloy type Hastelloy C276,austenitic stainless steels type 304NG,AL-6XN,HR3C.NF709 and SAVE 25,ferritic/martensitic(F/M) steel type P92,P122 and 410,and oxide dispersion strengthened steel MA 956,are tested.This paper presents corrosion rate,and focuses on the formation and breakdown of corrosion oxide film,and proposes the future trend for the development of SCWR internal structure materials.

Absorption effect of pure nickel on the corrosion behaviors of the GH3535 alloy in tellurium vapor

In this study,pure Ni was demonstrated to protect the GH3535 alloy from Te vapor corrosion because of its strong absorption capacity.Severe Te corrosion of a single GH3535 alloy sample occurred in Te vapor at 700C,which manifested as complex surface corrosion products and deep intergranular cracks.However,when pure Ni and the GH3535 alloy were put together in the vessel,the GH3535 alloy was completely protected from Te corrosion at the expense of the pure Ni.Thermodynamic calculations proved that the preferential reaction between pure Ni and Te vapor reduced the activity of Te vapor considerably,preventing the corrosion of the GH3535 alloy.Our study reveals one potential approach for protecting the alloys used in molten-salt reactors from Te corrosion.

The effect of Nb additive on Te-induced stress corrosion cracking in Ni alloy: a first-principles calculation

Nb can improve the resistance of Ni-based Hastelloy N alloy to Te-induced intergranular embrittlement.First-principles calculations are performed to research this mechanism by simulating the Ni(111) surface and the 5(012) grain boundary. The calculated adsorption energy suggests that Te atoms prefer diffusing along the grain boundary to forming the surface-reaction layer with Nb on surface of the Ni alloy. First-principles tensile tests show that the Nb segregation can enhance the cohesion of grain boundary. The strong Nb-Ni bonding can prevent the Te migration into the inside of the alloy. According to the Rice-Wang model, the strengthening/embrittling energies of Nb and Te are calculated, along with their mechanical and chemical components.The chemical bonds and electronic structures are analyzed to uncover the physical origin of the different effects of Te and Nb. Our work sheds lights on the effect of Nb additive on the Te-induced intergranular embrittlement in Hastelloy N alloy on the atomic and electronic level.

Synchrotron radiation-based materials characterization techniques shed light on molten salt reactor alloys

From a safety point of view, it is important to study the damages and reliability of molten salt reactor structural alloy materials, which are subjected to extreme environments due to neutron irradiation, molten salt corrosion, fission product attacks, thermal stress, and even combinations of these. In the past few years, synchrotron radiation-based materials characterization techniques have proven to be effective in revealing the microstructural evolution and failure mechanisms of the alloys under surrogating operation conditions. Here, we review the recent progress in the investigations of molten salt corrosion,tellurium(Te) corrosion, and alloy design. The valence states and distribution of chromium(Cr) atoms, and the diffusion and local atomic structure of Te atoms near the surface of corroded alloys have been investigated using synchrotron radiation techniques, which considerably deepen the understandings on the molten salt and Te corrosion behaviors. Furthermore, the structure and size distribution of the second phases in the alloys have been obtained, which are helpful for the future development of new alloy materials.

Microstructure Analysis for Chemical Interaction between Cesium and SUS316 Steel in Fast Breeder Reactor Application

The objective of this study is to presume cesium corrosion process and its dominant factors in SUS316 steel, a fuel cladding material for fast breeder reactor application, based on both experimental results of cesium corrosion out-pile test and thermodynamic consideration. The cesium corrosion test was performed in simulated environment of high burn-up fuel pin. And main corrosion products in the specimen after the corrosion test were specified by TEM （transition electron microscopy） and SEM （scanning electron microscopy） in order to formulate a hypothesis of the cesium corrosion process. At the end of this study, it was found that the dominant factors of the corrosion process are the amount of cesium on the surface of the specimen, chromium content in the alloy, the supply rate of oxygen and temperature.

堆内构件螺栓辐照促应力腐蚀可靠性评估方法

堆内构件连接螺栓所处工况环境恶劣,以辐照促进应力腐蚀开裂(IASCC)为代表的螺栓断裂失效事件时有发生,为了分析评估堆内构件螺栓的可靠性,提出了一种考虑螺栓IASCC失效的可靠性评估方法。首先,基于断裂力学理论中的应力强度因子准则,分别构建了SCC和IASCC失效的极限状态函数,并利用仿真分析方法求解函数中的最大等效应力。其次,考虑函数中各参数的不确定性,应用一次二阶矩法求解堆内构件连接螺栓的可靠度。最后,以堆内使用数量最多的M16型号螺栓为例进行计算分析。结果表明,断裂韧度与螺栓所受最大等效应力对螺栓可靠性的影响最大,裂纹扩展系数、裂纹扩展指数的影响次之,裂纹初始尺寸对可靠性的影响最小。此种评估方法可用于反应堆结构设计中各类连接螺栓的寿命预测及可靠性评估,也可以辅助确定反应堆结构安全检测周期和检测重点。

Characterization of solid particles sampled from condensates in boiling water reactor

The growth, activation and deposition of corrosion products are the primary sources of radiation buildup on the surface of out-of-core piping in nuclear power plants. The buildup of radiation can have negative effects on the performance of the facility and cause harm to staff during maintenance outages for refueling. This paper reports on the crystalline and amorphous structures of corrosion products sampled in the boiling water reactors in nuclear power plants of Kuo-Sheng and identified using an acid dissolving technique. X-ray diffraction, scanning electron microprobe and inductively coupled plasmaatomic emission spectroscopy were used to analyze the samples. The results indicate that the quantity of amorphous iron oxide at inlet of the condensate demineralizer in Unit 2 is higher than that in Unit 1. The proportion of crystalline to amorphous corrosion products can affect the efficiency of removal. Thus, these results can be used to explain the difference in removal efficiency of condensate demineralizers in different units. Moreover, the iron oxide structures with various properties were observed in different operational periods. It is probable that the higher proportion of amorphous structures with a smaller particle size would reduce efficiency in the removal of condensate demineralization in Unit 2.

表面完整性对制氢反应器材料耐腐蚀性影响

为研究某制氢反应器用钢表面完整性对其耐腐蚀性的影响,基于制氢反应器所产生的酸性介质腐蚀工况,对某制氢反应器用钢开展静态腐蚀试验。分别分析了不同表面粗糙度、不同表面残余应力对某制氢反应器用钢腐蚀速率、腐蚀后表面粗糙度与拉伸性能衰退的影响规律,并从晶间应力等方面分析表面完整性对耐腐蚀性的影响。通过试验发现:粗糙度较大的试样随腐蚀时间增加腐蚀速率降低,当表面粗糙度Ra为0.05μm时试样腐蚀速率随腐蚀时间增加而增大;腐蚀后粗糙度较大的试样表面粗糙度增大,而较小的试样表面粗糙度减小;腐蚀后材料的力学性能出现随表面粗糙度先增大后减小的趋势;随试件表面残余应力增大,腐蚀速率出现先减小后增大的趋势;腐蚀后表面粗糙度随初始残余应力的增大而增大;当初始表面残余应力为-293 MPa时,试样腐蚀后力学性能相对较好。

S Zorb装置反应器接收器脱气线和再生器取热盘管改造

某公司对S Zorb装置首个运行周期内出现的反应器接收器脱气线泄漏、再生器取热盘管泄漏等运行问题开展了技术攻关,提出了工艺优化调整、工程技术改造建议。2019年大检修时对反应器脱气线进行扩径,由DN100扩径为DN150,控制气提量小于670 m^(3)/h,对再生系统取热蒸汽和再生风流控阀等进行了工程技术改造。第二运行周期未出现脱气线减薄泄漏和再生器取热盘管泄漏问题。2023年6月大检修时对上次大检修改造部位进行检查以验证工程改造的有效性,经全面检查,确认脱气线内部未发现冲刷痕迹,再生器取热盘管打压合格,外观正常,证明改造措施有效,可以满足“四年一修”目标要求。

基于物理去除方法的反应堆及一回路系统腐蚀产物过滤系统设计

随着核反应堆服役年限的增加,反应堆一回路系统的设备、管道、阀门内表面存在一定的剥落,形成腐蚀产物悬浮于冷却剂中,最终变为活化腐蚀产物。随着活化腐蚀产物的不断增加会造成冷却剂中总固水平增加和反应堆厂房内空间剂量水平升高,也会对设备的运行造成一定的影响。为解决这一问题,本文采用在一回路系统上增设并联支路,支路中设置腐蚀产物过滤系统,采用物理去除的方法不断对冷却剂过滤去除其内的腐蚀产物,进而降低冷却剂中的总固水平。对于腐蚀产物过滤系统的过滤器,采用碟片式结果,采用有限元方法进行了过滤器单层固液两相分离仿真。经工程应用验证表明,本文设计的腐蚀产物过滤系统可有效降低冷却剂中的总固水平,表明对去污冷却剂中的腐蚀产物效果明显。

游泳池式轻水反应堆池壁池底及关键构件老化评估

核反应堆老化问题是决定反应堆安全运行的重要因素之一。中国原子能科学研究院游泳池式轻水反应堆(简称49-2堆)是一座典型的延寿运行的研究堆,也是中国目前运行时间最长的反应堆,开展老化研究对评估其继续延寿运行具有重要意义。本文主要以游泳池式轻水反应堆池底池壁(L03铝材)作为研究对象,对某些点缺陷的腐蚀情况进行分析,结果表明,49-2堆一次水水质条件引起铝合金的化学腐蚀较小,通过水下摄像机的检查,发现点缺陷无扩大和其他性质的变化,同时对池底地脚螺栓的检测,发现螺栓腐蚀程度轻微,X射线探伤表明内部无结构损伤情况,可侧面判定反应堆池壁池底的腐蚀情况较为良好,不会产生破口失水事故的发生。该结果对许可证延续申请具有重要参考价值。

反应堆合金材料腐蚀的分子动力学模拟研究进展

反应堆合金材料的腐蚀问题直接关系到反应堆的服役安全与寿命,一直以来都受到极大关注与广泛研究。但是仅依靠实验不足以明确其腐蚀机理,也难以高精度地预测其腐蚀行为。随着计算材料学的发展,模拟计算已经成为反应堆合金材料腐蚀研究的新手段。分子动力学方法可处理数万到数十万原子数量的尺度,适用于模拟多种材料的各种表面、界面行为,因此,近些年在反应堆合金材料腐蚀机理研究领域有了一定应用。本文首先回顾了分子动力学模拟方法,包括经典分子动力学方法、半经验分子动力学方法以及基于机器学习的分子动力学模拟。然后综述了反应堆合金材料腐蚀的分子动力学模拟研究方面的进展情况:一是适用于腐蚀模拟计算的分子动力学方法及势函数,主要包括反应力场、紧束缚量子化学力场和机器学习力场等;二是使用分子动力学方法研究水冷堆、液态金属冷却堆以及其他环境下反应堆合金材料的腐蚀研究现状介绍,包括晶界元素偏析、固液界面吸附和应力腐蚀开裂等腐蚀行为。

压水堆一回路腐蚀产物释放、迁移及沉积试验技术的研究进展

回顾了压水堆(PWR)腐蚀产物释放、迁移及沉积试验技术,介绍了基于试验技术获得的试验数据和结果,对比分析了不同试验技术的优缺点。结果表明:在腐蚀产物释放方面,基于定量水化学分析技术的腐蚀产物溶出释放试验可以在精确控制水化学条件的情况下直接检测腐蚀产物释放量,是一种高效、直接、精确的试验方法;在腐蚀产物迁移沉积方面,基于堆外动水回路的腐蚀产物沉积试验能够较好模拟压水堆实际工况,可实现性高且成本相对较低,试验条件可涵盖不同工况与设计范围,适于开展影响因素分析、规律探索和工程验证。目前,国内外获取的腐蚀产物特性数据较少,缺乏停堆工况下的试验研究,因此有必要系统开展不同工况下的腐蚀产物特性试验,获得一回路主要部件腐蚀产物释放、迁移及沉积试验数据,为国产源项软件验证提供支撑。

模拟压水堆一回路环境下冷应变对321不锈钢高温电化学行为和应力腐蚀开裂行为的影响

321不锈钢是常用的压水堆结构材料之一,在成型加工和服役期间易因各种因素发生冷应变,使其性能发生改变。本文在模拟压水堆一回路水化学环境中,测量了不同冷应变量321不锈钢的电化学阻抗谱,并用热应变样品作为对比;采用慢应变速率拉伸测试了冷应变试样应力腐蚀开裂性能。使用X射线衍射(XRD)、扫描电子显微镜(SEM)、能谱仪(EDS)对样品微观特征进行了分析。XRD分析表明,冷应变使基体发生了由奥氏体到马氏体的转变,而高温抑制了这一过程。随着应变程度的增大(至20%),电荷转移电阻增大,膜电阻随马氏体含量的升高而降低。裂纹萌生实验结果表明,马氏体优先发生氧化腐蚀,保护了奥氏体基体,抑制了应力腐蚀裂纹萌生。

反应堆合金材料腐蚀的密度泛函理论计算研究现状

经过几十年的研究,对现有反应堆合金材料服役工况下的腐蚀问题及规律已经有了较为清晰的认识,但是依然存在一些待解决的问题,如对反应堆合金材料腐蚀中本质过程的理解,单一因素在腐蚀过程中所起的作用,新材料以及极端环境下的腐蚀行为预测等。基于量子力学的密度泛函理论(Density Functional Theory,DFT)计算可以精确地预测极短时间内原子的运动过程以及相关能量的变化,近年来逐渐成为反应堆合金材料腐蚀过程研究的重要辅助手段,这将有助于解决上述问题。本文首先介绍了DFT方法,主要包括理论基础、发展过程与主流计算软件。然后对DFT理论在反应堆合金材料腐蚀领域应用的研究现状进行了综合论述和分析,包括水冷堆、液态金属冷却堆、熔盐堆介质环境中反应堆合金材料表面的吸附、解离、结合、内部扩散等腐蚀行为。最后展望了DFT应用在反应堆合金材料腐蚀方面的未来发展趋势。

一回路管道活化腐蚀产物放射性活度计算方法

对不同理论模型、模型方程及其优缺点等方面进行了综述,结合一回路活化腐蚀产物的迁移特性,推导出一种计算方法,对压水堆一回路管道活化腐蚀产物60Co、58Co的放射性活度进行了评估,通过Pactole计算结果对比论证了该方法的可靠性。

MAX相材料的耐铅铋腐蚀性能研究进展

铅冷快堆(LFR)以具有良好的热物理特性和化学稳定性的液态铅或铅铋合金(LBE)作为冷却剂,但是高温下LBE与结构材料长期接触带来的腐蚀问题会导致材料失效并带来安全隐患。MAX相材料因其优异的力学性能、高温耐腐蚀性能和抗辐照性能,成为铅冷快堆的重要结构材料之一。Ti-Si-C、Ti-Al-C和Zr-Al-C是目前被用于进行LBE腐蚀研究的三种常见MAX相材料体系。近年来,有越来越多的研究者们在不同腐蚀条件下的LBE中开展了腐蚀实验,但是对于MAX相材料的LBE腐蚀机理的分析仍不够系统和深入,也少有提高耐腐蚀性能的改性方法被提出。本文在简要介绍MAX相材料的结构、化学组成和制备方法之后,归纳综述了MAX相材料在高温液态LBE中的腐蚀行为及腐蚀机理的研究进展,提到了温度、氧浓度、时间和流速等不同因素对腐蚀层结构的影响。最后,探讨将用于提升MAX相材料在其他高温介质中耐腐蚀性能的改性方法,引入到针对耐LBE腐蚀性能研究的可能性。

基于氧化腐蚀行为的铅铋堆燃料组件多物理耦合特性研究

氧是铅铋堆中最具应用潜力的非金属缓蚀剂,在冷却剂中添加一定浓度的氧,可在结构材料表面生成保护性氧化膜,可以极大程度上缓解液态铅铋对结构材料的腐蚀。在铅铋堆中,氧化层的生长-去除行为受温度、氧浓度、冷却剂流速、时间等多种因素影响,同时氧化层的生长也改变了堆芯的热工水力特性和中子物理参数,因此,研究铅铋堆的氧化腐蚀场、热工水力场和中子物理场的耦合作用对铅铋堆应用有重要意义。本文基于MOOSE(面向对象的多物理场仿真环境)平台搭建了核-热-材多物理场耦合框架,开展了铅铋堆在基准工况下的核-热-材耦合分析,并研究了氧浓度和冷却剂入口温度对关键耦合参数时序变化规律和氧化层分布的影响。结果表明,基准工况下氧化腐蚀10 000 h后,燃料组件包壳表面的氧化层平均厚度约为9.86μm,燃料最大温升为13.36 K,k_(eff)下降7 pcm;氧浓度升高可以极有效地抑制磁铁矿溶解,但达到一定浓度后氧浓度的升高对Fe-Cr尖晶石的生长促进作用较小;冷却剂入口温度的升高会导致组件中心处包壳壁面的磁铁矿去除速率增大,并且可以大幅促进Fe-Cr尖晶石的生长。

基于氧化腐蚀特性的铅铋堆单棒核-热-材耦合研究

铅铋堆的氧化腐蚀受温度、氧浓度、时间等多因素影响,同时氧化层的生长也改变了堆芯热工水力和中子物理参数,因此,研究铅铋堆的氧化腐蚀场、热工水力场和中子物理场的耦合作用对铅铋堆应用有重要意义。本工作基于面向对象的多物理场仿真环境(Multiphysics Object-Ori1ented Simulation Environment,MOOSE)搭建了核-热-材多物理场耦合框架,开展了单棒在基准工况下的核-热-材耦合分析,并研究了氧浓度对关键耦合参数的影响。结果表明:基准工况下氧化腐蚀10000 h后氧化层平均厚度约为10μm,燃料最大温升为16 K,k_(inf)下降10^(-4);氧浓度升高可以极有效地抑制磁铁矿溶解,但是对Fe-Cr尖晶石生长的促进作用较小。