轻水堆核电站奥氏体不锈钢铸件的热老化及其老化管理

在轻水堆核电站中,奥氏体不锈钢铸件(CASS)在运行温度下长期工作将面临热老化问题。本文对热老化的机理进行了描述,并针对核电站的热老化问题,给出了老化管理工作的主要步骤,即部件敏感性甄别、老化状态评估和ISI大纲更新;结合CASS热老化的老化管理实践,对我国核电站的热老化管理工作提出了建议。     

316LN奥氏体不锈钢焊缝低温热老化行为研究

在压水堆核电厂中,主管道奥氏体不锈钢焊缝长期在其热老化敏感温度(280~325℃)下运行,为了研究主管道奥氏体不锈钢焊缝在核电厂运行温度下的热老化性能,开展了铁素体含量为10.7%的316LN不锈钢主管道焊缝在325、365、400℃下的低温热老化行为研究。结果表明:经6000 h热老化后,焊缝中铁素体相和奥氏体相中的主要元素含量没有发生明显变化,焊缝显微硬度快速增加但奥氏体相显微硬度没有发生变化,焊缝冲击功显著下降、拉伸性能变化较小。     

基于磁性多参数的铸造奥氏体不锈钢热老化状态评估研究

研究了应用磁性无损检测方法评估核电站主管道用铸造奥氏体不锈钢 (CASS)的热老化状态,结果表明在400 ℃下热老化不同时间,CASS的冲击功随着热老化时间延长逐渐降低,受材料组织不均应等因素影响,单一磁性参数与CASS的热老化时间之间无对应的单调规律;以磁性多参数测试为基础,结合主成分分析和非线性多元回归对测得的磁性...     

核电用铸造奥氏体不锈钢的热老化性能研究

压水堆核电厂主管道用国产铸造奥氏体不锈钢(CASS)长期服役会面临着热老化问题。在400℃高温下开展CASS材料的加速热老化实验,采用夏比冲击试验获得了材料室温吸收能量随热老化时间延长的下降规律,采用扫描电镜观察到不同热老化时间冲击断口形貌的变化趋势。实验结果表明:经过15 000h的加速热老化实验,CASS材料的热老化程度逐渐达到饱和状态,吸收能量虽大幅下降,但仍能满足设计规范对CASS材料在未老化状态时的考核要求。     

显微硬度法分析核电站主管道热老化趋势

使用显微维氏硬度计和冲击试验机研究了核电站主管道材料Z3CN20.09M在400 ℃加速热老化10 000 h前后的力学性能变化。结果表明,热老化导致试验材料的冲击吸收能下降;构成试验材料的铁素体相的显微维氏硬度上升,奥氏体相的显微维氏硬度基本保持不变。通过研究材料组织特征,剖析显微硬度与冲击韧性的关系,探索将显微硬度测试方法作为核电站主管道材料热老化趋势预测方法的可能性。     

中子辐照奥氏体不锈钢焊缝热影响区力学性能与微观结构的关系

奥氏体不锈钢通常用来制造核反应堆内部部件（如堆芯围筒）,经中子辐照后,其微观结构发生变化,从而导致力学性能的变化。焊缝热影响区材料在熔焊过程中受到焊接热循环的影响。本文研究了中子辐照对奥氏体不锈钢焊缝热影响区材料微观结构及力学性能的影响。试验材料选用退役压水堆中的AISI304不锈钢焊件。该材料在反应堆中经历了11次反应堆循环,最大辐照剂量为0．35dpa,温度为573K。通过试验得到不同辐照剂量下,热影响区材料和母材区材料的力学性能和微观结构。力学性能通过拉伸实验获得,采用室温和573K两种不同试验温度。用透射电镜观察材料的微观结构。运用弥散障碍硬化模型得到力学性能与微观结构之间的关系。试验结果表明：仅用透射电镜中观察到的中子辐照产生的缺陷并不能很好解释中子辐照硬化。在透射电镜中观察不到的那些小的辐照缺陷也能产生辐照硬化。     

Z3CN20.09M奥氏体不锈钢热老化冲击性能试验研究

采用GB/T19748-2005钢材夏比V型缺口摆锤冲击试验仪器化试验方法,对压水堆核电厂用离心铸造Z3CN20.09M奥氏体不锈钢主管道样品进行了实验室热老化的冲击性能研究。冲击试验数据的统计分析表明,热老化对Fiu/Fm比值不产生影响,而对冲击载荷有显著影响,对冲击能量的影响则更为显著。透射电子显微分析表明,热老化导致铁素体中出现沉淀物,并引发了奥氏体中位错组态的改变。与热老化时间lg t之间也满足线性关系。     

奥氏体不锈钢的显微组织及氦脆断口观察

利用电子显微镜对几种奥氏体不锈钢的形变和时效后的显微组织作了研究,在低应变下加工硬化主要依靠位错密度的增加,高应变下还出现大量形变诱发组织。时效引起碳化物的析出,基体还有再结晶的趋势。此外,还对经α粒子辐照后的氦脆断口作了观察。     

钼离子注入奥氏体不锈钢引发亚结构的变化

用加速电压为５０ｋＶ，剂量为２×１０＾１７ｉｏｎｓ／ｃｍ＾２的钼离子注入经１１５０℃固溶处理的奥氏体不锈钢中，利用透射电子显微镜弱束成像技术研究了被注入材料的横截面样品，发现离子辐照引发被注入亚结构变化的深度远大于离子本身的注入深度，其亚结构为多种位错组成的复杂位错组态，这种亚结构会使被注入材料具发于加工硬化的效果。探讨了离子注入材料中由辐射引发的加工硬化现象，即所谓的“长程效应”的原因。     

热老化对核电厂一回路管道失效概率的影响

建立基于热老化的管道失效概率计算流程。在实验研究的基础上,采取美国阿贡实验室的流程预测某管道材料在280、330℃下热老化后断裂韧性随运行时间的变化。计算含单个环向内表面裂纹的管道在考虑热老化与不考虑热老化2种情况下的累积失效概率。计算结果表明,考虑热老化因素得到的失效概率高于未考虑热老化的情况。在考虑热老化的情况下,较高的温度下热老化严重且管道失效概率更高。     

Fracture mechanics evaluation of cast duplex stainless steel after thermal aging

For the primary coolant piping of PWRs in Japan, cast duplex stainless steel, which is excellent in terms of strength, corrosion resistance and weldability, has conventionally been used. Cast duplex stainless steel contains the ferrite phase in the austenite matrix, and thermal aging after long-term service is known to decrease fracture toughness. Therefore, we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secure, even when such through-wall crack length is assumed to be as large as the fatigue crack length grown for a service period of up to 60 years.     

Influence of thermal aging on cast stainless steels used in JAEA's nuclear reactor Fugen

The influence of thermal aging on cast stainless steels used in JAEA's advanced thermal prototype reactor “Fugen” that were exposed to 275°C for about 15 years was investigated. The degree and mechanism of thermal embrittlement were evaluated on the basis of Charpy impact test results and the three-dimensional atom probe (3DAP) analysis of the microstructural changes in the spinodal decomposition using materials obtained from Fugen. The results revealed early signs of a thermal aging effect over a long period at the low temperature.     

1000MW核电管板纯净钢锻件制造工艺及其性能

根据1000MW核电管板用纯净钢锻件性能和组织的要求，利用合金化原理确定了冶炼时钢中各合金元素的成分控制方向；采用电炉加钢包炉加真空浇注进行冶炼浇注，真空浇注过程中加保护防止二次氧化；采用特殊的镦粗工艺避免了管板镦粗过程心部产生超标缺陷；采用合理的热处理工艺，保证管板锻件的组织和性能。经检验表明，锻件用钢的质量达到了纯净钢的要求，管板锻件的综合性能达到世界领先水平。     

Mechanical property and microstructural change by thermal aging of SCS14A cast duplex stainless steel

The aging behavior, especially saturation, of JIS SCS14A cast duplex stainless steels was investigated on the basis of the mechanical properties and microstructural changes during accelerated aging at 350 °C and 400 °C. The aging behavior of the materials mainly proceeds via two stages. During the first stage, the generation and concentration of the iron-rich and chromium-enriched phase in ferrite occurs by phase decomposition. The first stage corresponds to aging times of up to 3000 h at 400 °C. During the first stage, the ferrite hardness achieved is approximately 600 VHN, and the Charpy impact energy is almost saturated. During the second stage, the precipitated chromium-enriched phase aggregates and coarsens, and the G phase precipitation also occurs. The second stage corresponds to the aging times range of 3000-30 000 h at 400 °C. During the second stage, the ferrite hardness achieved is about 800 VHN; however, further hardening exceeding 600 VHN does not influence the Charpy impact energy.     

Assessment of thermal aging embrittlement in a cast stainless steel valve and its effect on the structural integrity

This paper analyzes the thermal aging embrittlement occurred in a cast stainless steel valve, which is part of the reactor water clean-up (RWCU) system of a Spanish boiling water reactor (BWR) nuclear power plant. The aim is to estimate the current and future state of the material and the corresponding structural integrity of the valve. Given that there is no data available for the experimental characterization of the material, the evolution of the mechanical properties (fracture toughness, yield stress, flow stress and Ramberg-Osgood parameters) has been estimated using the ANL procedure.With the obtained estimations, the critical crack size has been calculated using the European procedure FITNET FFS and the ASME Code.This analysis considers not only the evolution of the mechanical properties up to now but also its future evolution in case there is a life extension of the plant until year 2029.     

Aging degradation of cast stainless steel: Status and program

A program has been initiated to investigate the significance of in-service embrittlement of cast duplex stainless steels under light-water reactor operating conditions. The existing data are reviewed to determine the critical parameters that control the aging behavior and to define the objectives and scope of the investigation. The test matrices for microstructural studies and mechanical property measurements are presented. The initial experimental effort is focussed on characterizing the microstructure of long-term, low-temperature aged material. Specimens from three heats of cast CF-8 and CF-8M stainless steel aged for up to 70,000 h at 300, 350, and 400°C were obtained from George Fisher Ltd., of Switzerland. Initial analyses reveal the formation of three different types of precipitates which are not α′. An FCC phase, similar to the M₂₃C₆ precipitates, was present in all the long-term aged material.     

Effects of thermal aging on the low cycle fatigue behavior of austenitic–ferritic duplex cast stainless steel

Thermal aging is observed in a primary reactor cooling system (RCS) made of cast stainless steel. The observation occurs when the RCS is exposed for a long period of time to a reactor operating temperature between 290 and 330°C. An investigation of the effect of thermal aging on the low cycle fatigue characteristics of CF8M is required. The purpose of the present investigation is to find the effect of thermal aging of CF8M on a low cycle fatigue life. The specimen of CF8M is prepared by an artificially accelerated aging technique maintained for 300 and 1800 h at 430°C, respectively. The low cycle fatigue tests for the virgin and two aged specimens are performed at room temperature for the strain amplitudes (_ta) of 0.3, 0.5, 0.8, 1.0, 1.2 and 1.5%. In the experiment, it is found that the fatigue life is rapidly reduces with an increase of aging time. The experimental fatigue life estimation formula, between the virgin and two aged specimens are obtained.     

The effects of large concentrations of helium on the mechanical properties of neutron-irradiated stainless steel

Tensile and creep properties have been determined on specimens of type 316 stainless steel irradiated in the High Flux Isotope Reactor in the range 380 to 785°C. Irradiation of type 316 in this reactor partially simulates fusion reactor irradiation, with displacement damage levels up to 120 dpa and helium contents up to 6000 appm achieved in two years. Samples irradiated in the annealed condition to about 100 dpa and 4000 appm helium showed an increased yield strength between 350 and 600°C and, except at 350°C, a reduced ultimate tensile strength compared with values for the unirradiated material. Samples irradiated in the 20%-cold-worked condition showed decreases in both yield and ultimate tensile strengths at all test temperatures. The irradiated samples of both annealed and cold-worked material exhibited little strain hardening, and total elongations were small and became zero,for tests at 650° C. Tensile tests at 575°C and creep-rupture tests at 550°C showed strong effects of fluence on strength and ductility for helium contents above about 30 appm. Optical metallography showed extensive carbide precipitation at all temperatures and precipitation of a second phase, believed to be sigma, at the higher temperatures.     

The influence of helium on the high temperature mechanical properties of an austenitic stainless steel

Foil specimens of D1N 1.4970 austenitic stainless steel were cyclotron implanted to uniform helium concentrations up to 150 ppm. Subsequently, tensile and creep tests at different stress levels and temperatures between 600 and 800°C were performed and compared with the results from He-free control specimens. Whereas the latter exhibit transgranular ductile fracture, the implanted material breaks by intergranular failure with a corresponding reduction of the elongation to fracture values by factors between 3 and 6. These results and the TEM and SEM investigations of the microstructure are discussed by means of an embrittlement model which is based on the stress assisted nucleation and growth of helium filled grain boundary cavities.     

The evaluation of intergranular stress corrosion cracking problems of stainless steel piping in Taiwan BWR-6 nuclear power plant

Taiwan BWR-6 Kuosheng Nuclear Power Plant Unit 1 implemented the inspection of the intergranular stress corrosion cracking (IGSCC)-susceptible weldments of stainless steel piping in the reactor recirculation, reactor water clean-up, residual heat removal, core spray and feedwater systems. The purpose of this paper is to present the status of the fracture problems in the weldments. The crack growth analysis due to IGSCC and the standard weld overlay design based on the ASME Code Section XI and NUREG-0313 Rev. 2 for the fracture weldments are discussed in detail. Then, the contingent programs including the inspection program, fracture evaluation, and the standard weld overlay, are completely established to prevent pipe break during the reactor operation.