反应堆压力容器法兰密封面材料的性能研究

采用手工惰性气体钨极保护焊(TIG焊)制备反应堆压力容器密封面材料E308LMoT0-3、E309LMoT0-3及对比材料ER308 L的不锈钢堆焊层,对其进行硬度测试、显微组织观察、抗晶间腐蚀性能分析,以及点蚀点位测量和偏离水质条件下的局部腐蚀试验,研究E308LMoT0-3、E309LMoT0-3堆焊材料的点腐蚀和缝隙腐蚀性能以及腐蚀机理。试验结果表明,E308LMoT0-3、E309LMoT0-3焊丝堆焊后的试样除了具有良好的硬度、耐晶间腐蚀性能外,还具有良好的耐局部腐蚀性能,可以代替目前压水堆核电厂普遍使用的ER308L不锈钢堆焊材料。     

不锈钢堆焊层在Cl-溶液中的腐蚀性能研究

采用浸泡腐蚀试验方法,研究了不锈钢堆焊层材料在Cl^-溶液中的腐蚀情况,并通过金相显微镜、扫描电子显微镜、能谱分析观察表面形貌。研究表明,室温条件下堆焊层材料未发生任何腐蚀。在高温条件下,Cl^-的存在诱导了点腐蚀的发生,且随着Cl^-浓度的增加,点腐蚀加剧;较高浓度的Cl^-可导致缝隙内金属元素Cr的流失,缝隙腐蚀加深;应力腐蚀裂纹有沿晶开裂的特征,应力腐蚀敏感性随Cl^-浓度的增加有提高的趋势。     

误用四氯乙烯清洗剂对手动阀和密封组件应力腐蚀风险分析

某核电机组堆芯测量系统(RIC)手动阀和密封组件被误用四氯乙烯进行清洗。而该部件为奥氏体不锈钢,由于四氯乙烯热解过程中会导致氯离子Clˉ的产生,存在Clˉ致应力腐蚀风险。通过分析手动阀和密封组件的残留Clˉ含量、运行温度以及冷却剂中的氧含量,认为残留的Clˉ含量不会导致手动阀和密封组件在服役过程中发生应力腐蚀。     

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。      

国产690合金与321不锈钢异种金属焊缝的应力腐蚀行为研究

利用慢应变速率试验,采用非标准的漏斗状试样,对国产690合金与321不锈钢异种金属焊接部位(包括690合金热影响区、焊缝、321不锈钢热影响区)在100 mg/L Cl~(-1)除O_2条件下和100 mg/L Cl~(-1)饱和O_2条件下的应力腐蚀行为进行研究。并通过慢应变速率应力-位移曲线和断口形貌对微观组织、氯离子、氧含量对于材料的应力腐蚀(SCC)的影响进行分析。结果表明:690合金热影响区在100 mg/L Cl~(-1)除O_2条件下不易发生SCC,在100 mg/L Cl~(-1)饱和O_2条件下表现出一定的SCC倾向;321不锈钢热影响区在2种条件下均表现出明显的SCC倾向;690合金热影响区的粗大晶粒不利于塑性变形的晶粒间相互协调,导致了热影响区SCC的倾向增大。     

含草酸的硝酸溶液对316L不锈钢的腐蚀行为

模拟金属材料在草酸钚沉淀母液蒸发浓缩工艺的服役条件，按GB/T 4334.3-2000要求，开展了含草酸的硝酸溶液对316L不锈钢板材和焊件的腐蚀行为研究。实验采用称重法获得了腐蚀速率数据，采用扫描电镜观察金属表面的腐蚀形貌，通过能谱分析腐蚀前后样品的表面元素分布，并测定了腐蚀溶液中金属离子的浓度。结果表明：模拟实验条件下316L不锈钢各腐蚀样品腐蚀速率较低，为均匀腐蚀。在103 ℃、8.0 mol/L硝酸条件下，316L不锈钢板材和焊缝的平均腐蚀速率分别为0.049 6 g/(m²·h)和0.068 6 g/(m²·h)。研究结果为草酸钚沉淀母液蒸发浓缩设备的材料选择提供了数据参考，优化焊缝工艺是316L不锈钢作为草酸母液蒸发浓缩设备材料使用重要的改进方向。     

奥氏体不锈钢在MIPR模拟溶液中的腐蚀行为研究

研究了304L和321不锈钢在80℃、MIPR模拟溶液中的均匀腐蚀、晶间腐蚀和应力腐蚀行为.实验结果表明304L和321不锈钢腐蚀1 500 h后的表面腐蚀轻微,具有良好的耐均匀腐蚀性能,且无晶间腐蚀和应力腐蚀趋势.这是因为两种奥氏体不锈钢腐蚀后表面均形成了以Cr2O3为主的致密氧化膜,阻止了腐蚀的进行,表面Cr(Ⅲ)形成的外膜和内膜的协同作用提高了膜的稳定性和耐蚀性.     

液态锂铅合金中316L不锈钢的静态腐蚀行为

采用挂片法、失重法和金相分析,开展了结构材料316L不锈钢在液态锂铅(LiPb)合金中静态腐蚀行为的研究.研究结果表明:316L不锈钢中的组分元素,在液态LiPb合金中发生了溶解和质量迁移,这是导致材料腐蚀的主要原因,而温度和合金中的氧含量是影响静态腐蚀行为最重要的参数.     

304奥氏体不锈钢在不同浓度硝酸中的电化学行为

采用动电位极化曲线测量、开路电位测量等技术,研究了304奥氏体不锈钢在不同浓度硝酸溶液中的电化学腐蚀行为,并对304奥氏体不锈钢在硝酸溶液中的电化学反应历程进行了探讨。结果表明：304奥氏体不锈钢在硝酸溶液中具备不锈钢典型的极化曲线特征,有多个钝化区和过钝化区;硝酸浓度升高促进不锈钢表面钝化膜的生成,使开路电位向正电位方向移动,降低了硝酸溶液对不锈钢的腐蚀倾向,同时,随着硝酸浓度的升高,不锈钢的点蚀电位升高,提高了不锈钢耐点蚀能力;在硝酸溶液中,不锈钢的腐蚀速率同时受到酸度和硝酸根浓度的影响,二者相互矛盾,导致硝酸浓度对腐蚀速率的影响呈不规律性。结果表明,在0.5 mol/L硝酸中,不锈钢的腐蚀速率最高。     

冷变形316不锈钢在高温水中的应力腐蚀开裂行为

对不同冷变形量的核级316和316L不锈钢在高温水中的应力腐蚀开裂(SCC)行为进行了研究。通过试验,对溶解氧、氯离子和温度对裂纹扩展速率的影响进行了深入探讨和分析。试验结果显示,溶解氧和氯离子能明显加快材料的应力腐蚀开裂速率。当水化学条件一致时,325℃时的裂纹扩展速率较288℃时的裂纹扩展速率高。     

Effects of local radiolysis and geometric parameters on intergranular attack caused by crevice corrosion

Effects of γ-ray irradiation upon crevice corrosion (CC) of type 316L stainless steel (316L SS) as an initiation site of stress corrosion cracking in a boiling water reactor environment have been studied using a material corrosion test loop which could be irradiated with a ⁶⁰Co γ-ray source during testing. The CC tests were conducted using crevice specimens with various sizes of crevice gaps. Many of the examined specimen surfaces exhibited a selective grain boundary dissolution; that is, intergranular attack (IGA) as a result of the CC when the crevice gap size was lower than a certain value. The IGA initiation time was shortened by the γ-ray irradiation. The IGA occurred mostly near the crevice mouth at a distance of less than 2 mm from the mouth edge. When γ-ray exposure had occurred, it was found that the number of IGA sites deeper in the crevice increased compared with the IGA site distribution under the no-irradiation condition. Since the electrochemical corrosion potential inside crevice specimens must be low under the conditions for which IGA could occur, it was assumed that γ-ray irradiation accelerated the corrosion rate of 316L SS by decreasing the Fe²⁺ surface activity inside the crevice or increasing the cathodic current of radiolytic oxidants on the crevice surface. It was concluded that γ-ray irradiation affects the IGA occurrence not only temporally but also spatially.     

Evaluation of corrosion behaviors for reactor pressure vessels/primary containment vessels in Fukushima Daiichi units 1–3 nuclear power plant

Fukushima Daiichi nuclear power plants (1F) were damaged by unprecedented severe accident in the Great East Japan Earthquake on 11 March 2011, and seawater has been injected as an emergency countermeasure for the core cooling. Although, the RPV and PCV were not supposed to be exposed to diluted seawater, they have been exposed to diluted seawater environment or high-moisture environment. Therefore, seawater corrosion has become an important issue. Immersion corrosion tests were performed for low-alloy steel of RPV material and carbon steel of PCV material in 1F cooling-water-simulated environment. As a result, the mass loss by corrosion was reduced with the decreasing temperature and chloride ion concentration. Moreover, the effects of nitrogen deaeration and Na₂WO₄ addition on corrosion protection were remarkable among the selected corrosion countermeasures. In addition, the integrity assessments of RPV and PCV were performed considering the reduction of plate thickness based on corrosion test data and the load condition based on earthquake response analysis results. It had been confirmed that primary stresses for RPV and PCV equipment satisfied with the allowable values until at least 15 years after the accident.     

Modeling the effects of nuclear environments on crevice corrosion of pure nickel by coupling of phase and polarization behavior

As from long-term operating experience the high purity primary water cycle of light water nuclear reactors may exhibit excursions from the recommended water chemistry leading to potentially favorite conditions for stress corrosion cracking (SCC) which may be initiated and its propagation controlled by local pitting and crevice corrosion. Deterministic modeling of local corrosion including incubation times for crevice corrosion should therefore provide a basis for lifetime predictions of components, which have been subjected to sporadic intermediate water chemistry fluctuations. Based on previous work for room temperature (RT), the chloride-induced crevice corrosion at 288 °C of pure nickel as an important base element in respective high alloyed nuclear materials is modeled by coupling anodic polarization with the precipitation of nickel oxide and nickel chloride calculated from the water–hydrogen–nickel chloride heterogeneous phase equilibrium diagram. The surface corrosion potentials are fixed by bulk levels of hydrogen and oxygen contents as well as pH simulating hydrogen treatment of irradiation subjected cooling water for the reduction of corrosion potentials and mitigation of SCC at operating temperature 288 °C in Boiling Water Reactors (BWRs). Assuming chemical equilibrium conditions during the selected time steps in a relevant component crevice the calculated change of the crevice solution composition is quantitatively shown to initiate crevice corrosion by the breakdown of the passive nickel oxide layer followed by the formation of non-passive nickel chloride and the subsequent acidification of the crevice solution. The effects of corrosion potentials, bulk levels of pH and chlorides, are investigated. As a result, the reduction of corrosion potentials and increase in bulk pH provide significant increases in the passive layer breakdown times and acidification times inside the crevice. Depending on bulk pH and corrosion potentials the reduction of bulk chlorides down to recommended levels in BWRs retards crevice corrosion significantly. For a standard 100,000 h time for crevice acidification to locally less than pH = 0 the respective chloride–pH domain is evaluated. Such diagrams may be related to respective effects on stress corrosion cracking and its mitigation by hydrogen water chemistry (HWC).     

Characterization of the stress corrosion cracking behavior of thermally sensitized 20Cr-25Ni stainless steel in a simulated cooling pond environment

Niobium stabilized 20Cr-25Ni stainless steel is used for nuclear fuel cladding in the UK's fleet of advanced gas cooled reactors (AGRs). The cladding can have chromium-depleted grain boundaries as a consequence of irradiation in a reactor core, rendering a small proportion of cladding susceptible to intergranular stress corrosion cracking in cooling pond waters after removal from the reactor. In this work, thermal sensitization was used to simulate chromium depletion and the sensitized material was assessed for its susceptibility to pitting corrosion and stress corrosion cracking using slow strain rate testing (SSRT). Elevated chloride concentrations were used to accelerate corrosion initiation and propagation. In 10 ppm chloride and 80 °C, the pitting potential was at potentials between +375 mV and +400 mV (SCE). SSRT appeared to lower the pitting potential, with intergranular corrosion and intergranular stress corrosion cracks observed to nucleate at potentials of +200 mV (SCE).     

Evaluation method of corrosion lifetime of conventional stainless steel canister under oceanic air environment

Natural exposure and accelerated corrosion tests of conventional stainless steels for canisters of Types 304, 304L, and 316(LN) for concrete casks were conducted using several test specimens and 1/5 scale canister models. The welding residual stress of a full-scale model canister was also measured and the lifetime of sealability of canisters against corrosion evaluated. The maximum pitting rate and crevice corrosion rate of Type 304 were approximately 20 and 30 μm/year. Many SCC in the 4 Point Bending (4PB) test specimens were found to initiate from the bottom of the corrosion area by pitting or crevice corrosion. The SCC propagation rates in Types 304 and 304L under natural conditions were around 1.2E−12 to 1.8E−11 m/s in the K (Stress Intensity Factor) range of 0.6–9.0 MPa m^1/2, and that of the accelerated test (60 °C, 95% RHS, filled with NaCl mist) around 1.0E−10 to 3.5E−9 m/s in the K range of 0.5–30 MPa m^1/2. The SCC propagation rates under both natural and accelerated conditions were independent of K. The lifetime of sealability estimated from 1/5 scale models was longer than that from the small bending test specimens and has a safety margin as a structure.     

Probabilistic methodology to estimate environmental conditions for localized corrosion and stress corrosion cracking of Alloy 22 in a high-level radioactive waste repository setting

The US Department of Energy (DOE) has indicated that it may use Alloy 22 (Ni-22Cr-13Mo-4Fe-3W) as the waste package outer container material for the potential high-level waste repository at Yucca Mountain, Nevada. This alloy could be susceptible to localized corrosion, in the form of crevice corrosion, and stress corrosion cracking if environmental conditions and material requirements (e.g., existence of crevices or high enough tensile stresses) are met. An approach is proposed to assess the likelihood of environmental conditions capable of inducing crevice corrosion or stress corrosion cracking in Alloy 22. The approach is based on thermodynamic simulations of evaporation of porewaters and published equations to compute corrosion potential and critical potentials for crevice corrosion and stress corrosion cracking as functions of pH, ionic concentration, temperature, and metallurgical states from fabrication processes. Examples are presented to show how the approach can be used in system-level assessment of repository performance.     

Effect of magnesium on the lead induced corrosion and SCC of alloy 800 in neutral crevice solution at high temperature

Dissolved magnesium species in the feed water reduce the incidence of lead-induced stress corrosion cracking (PbSCC) of Alloy 800. The passivity of material was improved by replacing a part of chlorides in the lead-contaminated chemistry with magnesium chloride, as indicated by: (1) a higher pitting potential; (2) lower passive current densities; (3) a film structure with less defects and more spinel oxides. According to the constant extension rate tensile (CERT) tests conducted in the neutral crevice solutions at 300 °C, lead contamination would reduce the ultimate tensile strength (UTS) and elongation of material. The CERT test results were in agreement with the fracture morphology observations. Magnesium addition significantly reduced the detrimental effect of lead contamination.     

国产A508-3钢在模拟AP1000一回路水环境下的疲劳性能研究

反应堆压力容器（RPV）钢在一回路水环境下的疲劳性能是评价其设计寿命的重要参数。本文针对国产A508-3钢开展了模拟AP1000一回路水环境的低周疲劳性能试验研究，获得了321 ℃、155 MPa及01 ppm溶解氧水环境下的疲劳行为数据和断裂机理。研究结果表明，国产A508 3钢峰值应力随应变幅的增大而逐渐增大，疲劳试验过程中试样表现出循环硬化、循环软化和饱和3个阶段；在应变幅由02%逐渐增加至06%的过程中，疲劳周次从105逐渐降低至102；疲劳断口具有疲劳和腐蚀特征，属于典型的腐蚀疲劳断裂。     

Intergranular stress corrosion cracking behavior of niobium-added Type 308 stainless steel weld overlay metal in a simulated BWR environment

Type 308 stainless steel weld metal as an internal cladding of reactor pressure vessels for boiling water reactors is subject to postweld heat treatment during fabrication and can suffer sensitization depending on carbon and ferrite contents. This sensitization can be avoided by using niobium-added Type 308 weld metal (specified as Type 308 NbL) which was developed for one-layer overlay welding application. In the present study, stress corrosion cracking (SCC) behavior of heat-treated Types 308 and 308NbL weld metals in oxygenated high temperature pure water was evaluated by slow strain rate test and U-bend tests with and without crevice. Every test showed that Type 308NbL weld metals were highly resistant to SCC compared to ordinary Type 308 weld metals. In single U-bend test, one-layer overlay weld metals of Type 308NbL produced by electroslag welding process using wide strip electrodes were crack free over 23,000 h. The U-bend test data of ordinary Type 308 weld metals were successfully analyzed by an SCC reaction model. Using this analysis, the SCC life margin for Type 308NbL over ordinary Type 308 weld metals, expressed as a ratio of respective times to SCC initiation, was estimated to be about 36.