【摘要】 采用慢应变速率拉伸、U型弯曲和紧凑拉伸试验，模拟压水堆核电站一回路高温高压水环境，测试Inconel 690合金超音频脉冲方波电弧（Ultrasonic Frequency Pulsed Square Wave Arc,UFP）钨极氩弧焊堆焊层应力腐蚀性能，并与恒流（Constant Direct Current,DC）钨极氩弧焊堆焊层进行比较。结果发现，慢应变速率拉伸试验中，UFP试样和DC试样均未产生局部不连续的脆性开裂迹象。U型弯曲试验中，2个DC试样弯曲后浸泡前存在裂纹，浸泡后，该2个试样上产生多处裂纹，最长分别为304μm和550μm，另一个DC试样和全部3个UFP试样浸泡前无缺陷，浸泡后表面平整、完好，未发现腐蚀损伤和其他缺陷。紧凑拉伸试验中，不同取样方向试样、不同溶解氢含量时裂纹扩展速率不同，DC试样和UFP试样平均裂纹扩展速率分别为（1.766～2.088）×10^-3 mm/y和（2.372～2.889）×10^-3 mm/y。若以核电机组设计寿命60年计，则服役期内裂纹扩展长度分别为0.11～0.13 mm和0.14～0.17 mm。综合裂纹萌生与扩展试验结果，认为Inconel 690合金UFP-TIG工艺堆焊层在模拟压水堆一回路水中应力腐蚀敏感性极低。更多还原

【Abstract】 Slow Strain Rate Tensile（SSRT）, U-bend and Compact Tension（CT） tests were conducted to simulate the hightemperature and high-pressure water environment of a pressurized water reactor（PWR） nuclear power plant. The stress corrosion properties of the Inconel 690 alloy overlay welds fabricated using ultrasonic frequency pulsed square wave arc（UFP）tungsten inert gas（TIG） welding were evaluated and compared with those of constant direct current（DC） TIG welding overlays. The results showed that neither the UFP nor the DC specimens exhibited signs of localized discontinuous brittle cracking in the SSRT tests. In the U-bend tests, two DC specimens exhibited cracks prior to immersion, and after immersion,these two specimens developed multiple cracks with lengths of 304 μm and 550 μm, respectively. The other DC specimen and all three UFP specimens showed no defects before immersion and remained smooth and intact without any signs of corrosion damage and defects after immersion. In the CT tests, the crack propagation rates varied with different sampling directions and hydrogen contents. The average crack propagation rates for the DC and UFP specimens were（1.766-2.088）×10^-3mm/y and（2.372-2.889） ×10^-3 mm/y, respectively, corresponding to crack propagation lengths of 0.11-0.13 mm and 0.14-0.17) mm after 60 years of design life. Based on the comprehensive results of SCC initiation and propagation tests, it was concluded that the Inconel 690 alloy UFP-TIG overlay welds exhibits negligible SCC susceptibility in the simulated PWR primary water environment.更多还原