主管道用材中低加载速率下常温断裂性能研究

核电厂主管道在长期服役下,母材及焊接处可能出现裂纹,需要对主管道材料及焊接材料在中低加载速率下的断裂性能进行研究,避免主管道在强地震冲击下可能出现的双端剪切断裂。基于Instron VHS高速材料试验机,开发了一套材料在中低加载速率下的断裂性能测试方法,测量了核电厂主管道材料控氮00Cr17Ni12Mo2及焊接材料OK Tigrod 316L在0.5 m/s加载速率以内的常温断裂性能。结果表明,常温下核电厂主管道材料控氮00Cr17Ni12Mo2在0.5 m/s冲击速率以内并不启裂,焊接材料OK Tigrod 316L在0.5 m/s加载速率以内的断裂韧性并未出现明显的规律性变化。      

316不锈钢室温和350℃低周疲劳性能研究

采用MTS材料试验机研究作为反应堆结构材料的316奥氏体不锈钢母材在350℃和室温,以及焊缝在室温,±0.3%~1.5%应变幅的低周疲劳性能试验,并采用扫描电镜对试验后样品进行了断口分析。研究结果表明,316不锈钢疲劳性能较好,室温下疲劳寿命高出350℃同一应变幅的30%~50%以上,且母材的疲劳寿命显著高出焊缝同一应变幅的一倍以上。随应变幅的增加,材料疲劳寿命相应下降,峰值应力增加。室温下母材和焊缝均呈现出随循环周次增加、峰值随应力逐渐下降的规律。母材在高温下,随应变幅的增加,逐渐由循环硬化过渡到饱和行为。低周疲劳试验后,断口表面可观察到裂纹源和疲劳条带。随应变幅增加,疲劳条带间距增大,且同一应变幅下,焊缝的间距大于母材,高温的疲劳间距大于室温,与疲劳试验结果相吻合。     

不同材料的核电厂主管道LBB评估对比研究

我国核电厂主管道的材料主要有铸造奥氏体不锈钢(CASS)和锻造不锈钢(WSS)。针对CASS和WSS两种材料的主管道,依据美国核管会的SRP3.6.3进行主管道硬前漏(LBB)评估的对比研究。考虑热老化效应获取可信的材料性能数据,根据材料性能差异采用极限载荷法或J积分撕裂模量汇交法计算临界裂纹尺寸。根据Henry均匀非平衡双相流模型计算泄漏裂纹尺寸,并通过环向表面裂纹和贯穿裂纹的扩展分析论证了裂纹疲劳扩展不会导致管道的突然断裂。研究结果表明,WSS材料和CASS材料相比具有更好的LBB性能。     

动态载荷下奥氏体不锈钢管道LBB环向贯穿裂纹稳定性实验及理论分析方法研究

管道环向贯穿裂纹是否稳定是评判管道是否满足破前漏（LBB）设计准则的标准之一,为确保LBB技术安全可靠,对管道环向贯穿裂纹在动态载荷下的稳定性进行实验研究。采用水平冲击机对含环向贯穿裂纹的管道依次进行加载速度为1.22、2、3、4 m/s的高温不带运行压力的冲击实验,以获得各应变率下的实验极限载荷值,并与工程理论分析计算结果进行比较。分析表明:奥氏体不锈钢管道环向贯穿裂纹在动态载荷下的失效模式为塑性失稳;经实验验证,在工程中对承受动态载荷的奥氏体不锈钢管道进行LBB分析时,采用美国核管会标准审查大纲3.6.3破前漏评估程序（SRP 3.6.3）中的极限载荷理论分析方法具有较高的工程安全性,若同时选用准静态下的材料力学性能,则工程安全性更高。      

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。     

压水堆核电厂主管道窄间隙自动焊焊丝研究

根据窄间隙自动焊工艺及主管道母材特点,在ER316L焊材基础上研究与主管道自动焊技术相匹配的专用焊材。通过模拟焊接试验和热裂纹试验验证了自动焊焊材的稳定性、可焊性,并对其焊缝疲劳寿命进行试验。研究结果证明新开发的自动焊焊材与主管道窄间隙自动焊工艺相匹配,焊缝接头综合性能良好。     

核级不锈钢管座焊接区与母材区疲劳裂纹扩展性能对比研究

焊接接头广泛应用于核电站管座处,而疲劳裂纹扩展是导致焊接接头失效的重要原因之一。因此,研究焊接区材料的疲劳裂纹扩展和寿命预测方法对准确预测焊接接头的寿命具有重要意义。本文以核电厂常用的304L不锈钢焊缝材料为对象,研究不同载荷比、不同取样方向对疲劳裂纹扩展速率的影响;基于试验数据建立焊缝材料的疲劳裂纹扩展速率模型,并与美国机械工程师协会（ASME）标准中奥氏体钢进行对比。结果表明:不同取样方向对焊缝疲劳裂纹扩展速率的影响不大,但载荷比对其有较大影响,较低载荷比下,焊缝的疲劳裂纹扩展速率在某个应力强度因子幅值（?K）前高于母材的疲劳裂纹扩展速率,在其之后则低于母材,而较高载荷比下则恰恰相反。      

核级管道中环向贯穿裂纹的断裂韧性计算研究

裂纹稳定性分析是破前漏技术(LBB)的关键问题之一。在工程设计中,失效评定图法(FAD)广泛应用于管道的临界裂纹长度计算。文章以某核电厂主蒸汽管道材料SA335为研究对象,结合有限元方法与材料损伤模型模拟了SA355管道中环向贯穿裂纹的断裂过程,并得到其断裂韧性K_(IC)。计算结果发现,通过紧凑拉伸试样测量的K_(IC)低估了管道裂纹的断裂韧性,使得评定点坐标K_r容易落在评定曲线包络的区域之外,进而得到较为保守的临界裂纹长度。采用文中推荐的计算方法,可以合理并有效地评价核级管道中贯穿裂纹的裂纹稳定性。     

国产快堆包壳管材料——冷加工316（Ti）不锈钢高温强度下降的微观机理分析

国产快堆包壳管材料316（Ti）不锈钢（1995—1996年，由原上海第五钢厂研制）的力学性能测试结果表明：材料拉伸强度与国外材料相当，但高温蠕变和高温持久性能却低于国外的数据。将这种材料与俄罗斯快堆包壳材料CH-68比较，国产材料在室温和600℃下的屈服极限强度高于俄罗斯材料的，但625和700℃的高温持久强度却低于俄罗斯CH-68数据，尤其是在700℃下，国产材料的持久断裂强度不但大大低于俄罗斯材料，且强度下降很快。     

焊接残余应力对316LN不锈钢应力腐蚀裂纹扩展速率的影响

核电站不锈钢管道焊接过程中引入的残余应力对焊接接头的应力腐蚀开裂性能有较大影响。本文针对一AP1000主管道316LN不锈钢焊接模拟件进行残余应力分析和应力腐蚀裂纹扩展速率测量,得到了焊后原始状态和去应力热处理状态的焊接热影响区材料在高温高压水中的应力腐蚀裂纹扩展速率。实验结果表明,焊接残余应力明显提高了热影响区的应力腐蚀裂纹扩展速率,且在含氢的压水堆一回路正常水化学下焊接残余应力的影响更加显著。     

沉淀硬化马氏体不锈钢热老化脆化的热电势检测研究

研究了17-4PH沉淀硬化马氏体不锈钢在400℃下不同时效时间力学性能和热电势变化规律,提出了热电势和冲击韧性的经验公式,并通过在核电厂服役13 a的主蒸汽隔离阀阀杆进行验证。验证结果显示,随着时效时间的延长,材料的冲击韧性下降,屈服强度、抗拉强度和硬度升高,断面收缩率和断裂伸长率下降。材料的热电势变化与冲击韧性呈现指数相关性,材料的屈服强度、抗拉强度、硬度和热电势呈现出较好的线性关系。通过热电势检测评估的冲击韧性和实测值显示出较好的符合性。      

Effect of thermal aging on fracture toughness of RPV steel

The effect of thermal aging on mechanical properties and fracture toughness was investigated on pressure vessel steel of light water reactors. Submerged are welded plates of ASME SA508 C1.3 steel were isothermally aged at 350°C, 400°C and 450°C for up to 10,000 hrs. Tensile, Charpy impact and fracture toughness testings were conducted on the base metal and the weld heat affected zone (HAZ) material to evaluate whether thermal aging induced by the plant operation is critical for the integrity of the pressure vessel or not. Tensile properties of the base metal was not changed by thermal aging as far as the thermal aging conditions were concerned. Relatively distinct degradation was observed in fracture toughness J_IC and J-resistance properties of both the base metal and the weld HAZ material, while only slight changes were observed in Charpy impact properties for both of them. However, it was concluded that the effect of thermal aging estimated by 40–80 years of plant operation on fracture toughness of both materials is small.     

动态应变时效对18MND5低合金钢材料抗拉性能的影响分析及服役安全评价

18MND5低合金钢凭借其良好的机械性能,被广泛应用于压水堆核电站核岛设备上。针对核岛设备用国产18MND5低合金钢钢板,在性能热处理状态和性能热处理+模拟消应力热处理两种状态下,进行了室温及100、125、150、175、200、250、300、350 ℃拉伸试验。基于上述试验结果,分析了动态应变时效对不同热处理条件下18MND5低合金钢材料抗拉性能的影响。结果表明,在室温～350 ℃温度范围内,随着温度的升高,18MND5低合金钢材料的抗拉强度呈先下降后上升的特征,抗拉强度谷值出现在150～200 ℃,发生了动态应变时效。采用两种热循环制度模拟18MND5低合金钢材料的封头成形工艺,分析了热循环制度对材料抗拉性能的影响。结果表明,18MND5低合金钢材料150 ℃下的抗拉强度仍低于350 ℃下的抗拉强度,动态应变时效的影响仍存在;经历热循环后18MND5低合金钢材料的抗拉强度未低于原始材料性能态的抗拉强度。此外,根据RCC-M中设备设计用抗拉强度S_u的计算公式,对18MND5低合金钢材料在抗拉强度低谷的服役安全性进行了评价。结果表明,在动态应变时效影响下,18MND5低合金钢材料产生的抗拉强度的谷值仍处于材料安全使用范围内。     

基于卡尔曼滤波的热老化性能可靠性预测

以核电站主管道为研究对象,运用性能退化可靠性理论,对主管道的热老化性能可靠性进行了研究。首先通过加速热老化实验获得的数据,分析主管道奥氏体不锈钢材料冲击性能及断裂韧性的退化过程,利用状态空间方法建立了时变性能退化量模型,并通过卡尔曼滤波对性能趋势进行预测;然后考虑冲击性能与断裂韧性之间的相关性,运用随机过程理论建立了基于多性能参数的主管道热老化实时性能可靠性预测模型,从而得到多参数下的主管道热老化性能可靠度及可靠性寿命,为核电站进行主管道老化维修决策优化管理提供了科学依据。     

State of experience in the application of the steel grade 15 MnNi 6 3 for containments and components of nuclear power stations

Steel grade TSB 370/15 MnNi 6 3 presented first in 1976 as material for the erection of nuclear safety containments has been accepted very quickly in practice. The material of which yield point and high tensile values are in the range of the well-known steel StE 36, shows high toughness properties. This is valid either for the unaffected base material as well as for material conditions produced by fabrication as for instance in the HAZ formed by welding and after stress-relief annealing. There is no sign to expect crack formation in the HAZ after stress-relieving also in case of unfavourable process stress conditions. By means of wide plate tests with weld seams showing artificially introduced defects, the high safety in brittle fracture of this material in the range of the application temperatures could be shown. The concept of transition temperature (RT-NDT concept) as specified in the ASME rules and the German rules for materials for the construction of nuclear power stations can be met with great safety by application of material TSB 370/15 MnNi 6 3. The favourable properties of the base material and its good behaviour during fabrication have led to the fact that steel grade TSB 370/15 MnNi 6 3 has a constant place in the election of material for nuclear power station components.     

Toughness behaviour of the steel X 20 CrMoV 12 1 — Results of optimizing the base- and weld metal

The 12% Cr steels are frequently used in German power plants for tubings, pipes, rotors, and blades. The maximum operating temperature is limited by their creep strength properties to about 550°C. There are applications at even higher temperatures. Sufficient materials toughness is required for the base metal and weld metal to withstand sudden load changes. This is of special interest for use in nuclear power plants. Under operating conditions at elevated temperatures microstructural changes occur which greatly influence the toughness properties of both base metal and weld metal. This paper presents the results of ageing treatments at 550°C, carried out with a 12% Cr steel (DIN X 20 CrMoV 12 1) specifically optimized for toughness. The decrease in toughness is already evident at ageing times as low as 1000 h for conventional and optimized material. This drop in toughness is tentatively explained by differences in grain sizes and carbide content (M₂₃C₆ carbides). Detailed investigations indicate that additional carbide precipitation may significantly contribute to the decrease in toughness.     

Fracture properties evaluation of carbon steel piping for main steam line

The objective of this paper is to evaluate the material properties of SA106 Gr.C carbon steel pipes and their associated welds manufactured for main steam line of Yonggwang nuclear generating station units 3 and 4.Tensile and fracture toughness tests were performed on specimens taken from actual pipes and the effect of various parameters such as the pipe size, the specimen orientation, the test temperature and the welding procedure on the material properties are discussed. Fracture toughness test results show significant crack plane orientation and test temperature dependence. An empirical equation correlating tensile and fracture toughness data is also proposed.     

Fracture toughness of narrow-gap welded joints in the nuclear pressure vessel steel 22 NiMoCr 37

The mechanical testing of narrow-gap welded joints in 100 and 200 mm thick sections of the steel 22 NiMoCr 37 has revealed that the weld metal, and not the heat affected zone (HAZ) or the weld metal-parent metal boundary. is the critical region. This modified gas-shielded welding process operates with a very low heat input of the order of 6.500 J cm⁻¹ pass⁻¹ and the combination of small diameter welding wires and high welding speeds contributes to the excellent joint properties in the as-welded condition.To investigate the effect of preheating and post-welding heat treatment on the mechanical properties of narrow-gap welds, tensile, notch impact, flat bend and fracture toughness test specimens were extracted from joints welded with the following conditions: (1) no preheating: no post-weld heat treatment; (2) no preheating: soaking at 300°C: (3) no preheating: stress-relief heat treatment at 600°C; (4) preheating 200–250°C; no post-weld heat treatment; (5) preheating 200–250°C; soaking at 300°C; (6) preheating 200–250°C; stress relief heat treatment at 600°C. Tensile testing at room temperature and at 250°C of round specimens oriented across the seam revealed the ultimate fracture to be always located in the base material remote from the welded zone. Although pores or slag inclusions had an influence on bend-test results of specimens in the as-welded condition, the results generally show failure free bends to 180°C with no evidence of cracking in the HAZ or at the fusion boundary.Using sharp-notched impact bend specimens with the notch located in the centre of the seam as well as in and across the HAZ, absorbed energy-test temperature curves have been determined for each welding condition. In comparison with the base material impact toughness, the weld exhibits superior toughness in the temperature range − 60 – 0°C, but yielded lower values at room temperature. After stress relieving at 600°C, the impact toughness of the weld reduced significantly, apparently due to precipitations occurring in the weld-metal microstructure. Test results from welded specimens with the no notch in the HAZ show this region to have superior notch impact toughness to the base material.Crack opening displacement (COD) specimens 45 × 90 × 380 mm with the fatigue crack located in the weld metal and in the HAZ were tested at 0 and 20°C using both the recommendation in BS DD 19: 1972 as well as acoustic emission measurements for the determination of COD values. For this method of fracture toughness testing it has been shown that the occurrence of a critical event must be clearly defined as corresponding to stable crack growth or alternatively to unstable crack propagation.     

基于主曲线法的核电主管道材料断裂韧性研究

核电工程的防脆断设计和在役缺陷评价主要应用线弹性断裂力学,并基于材料断裂韧性进行评价。材料断裂韧性需通过试验测定,首先采用落锤试验和V型缺口冲击试验共同确定参考温度,或采用主曲线法确定参考温度,然后将参考温度和材料温度作为变量建立关系式描述材料的断裂韧性。主曲线法能通过较少的试样试验得到材料的断裂韧性,并具有较高的置信度,因此在工程中已得到越来越多的应用。文中采用ASTM E1921标准,应用主曲线法测量了某核电厂主管道材料的参考温度,确定了材料的断裂韧性,并与ASME第Ⅺ卷附录G中的断裂韧性进行比较。结果表明,采用主曲线法得到的材料断裂韧性更高,工程应用中减少了保守裕度,提高了经济性。