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首页> 外文会议>20th International Symposium on Effects of Radiation on Materials, Jun 6-8, 2000, Williamsburg, Virginia >Investigation of Temper Embrittlement in Reactor Pressure Vessel Steels Following Thermal Aging, Irradiation, and Thermal Annealing
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Investigation of Temper Embrittlement in Reactor Pressure Vessel Steels Following Thermal Aging, Irradiation, and Thermal Annealing

机译：热老化，辐照和热退火后反应堆压力容器钢的回火脆化研究

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The Heavy-Section Steel Irradiation Program at Oak Ridge National Laboratory includes a task to investigate the propensity for temper embrittlement in coarse grain regions of heat-affected zones in prototypic reactor pressure vessel (RPV) steel weldments as a consequence of irradiation and thermal annealing. For the present studies, five prototypic RPV steels with specifications of A302 grade B, A302 grade B (modified), A533 grade B class 1, and A508 class 2 were given two different austenitization treatments and various thermal aging treatments. Thermal aging treatments were conducted at 399, 425, 454 and 490℃ for times of 168 and 2000 h. Charpy V-notch impact toughness vs temperature curves were developed for each condition with ductile-brittle transition temperatures used as the basis for comparing the effects of the various heat treatments. Very high austenitization heat treatment produced extremely large grains which exhibited a very high propensity for temper embrittlement following thermal aging. Mergranular fracture was the predominant mode of failure in many of the materials and Auger analysis confirmed significant segregation of phosphorus at the grain boundaries. Lower temperature austenitization treatment performed in a super Gleeble to simulate prototypic coarse grain microstructures in submerged-arc weldments produced the expected grain size with varying propensity for temper embrittlement dependent on the material as well as on the thermal aging temperature and time. Although the lower temperature treatment resulted in decreased propensity for temper embrittlement, the results did provide motivation for the investigation of the potential for phosphorus segregation as a consequence of neutron irradiation and post-irradiation thermal annealing at 454℃. One of the A 302 grade B (modified) steels was given the Gleeble treatment, irradiated at 288℃ to about 0.8 X 10~(19)n/cm (>1 MeV) and given a thermal annealing treatment at 454℃ for 168 h. Charpy impact testing was conducted on the material in both the irradiated and irradiated/annealed conditions, as well as in the as-received condition. The results show that, although the material exhibited a relatively small Charpy impact 41-J temperature shift, the heat-affected zone-simulated material did exhibit significant intergranular fracture in the post-irradiation annealed condition.
机译：橡树岭国家实验室的重截面钢辐照计划包括一项任务，目的是研究由于辐照和热退火而导致原型反应堆压力容器（RPV）钢焊件中热影响区的粗晶粒区域的回火脆化倾向。对于本研究，对五种规格分别为A302 B级，A302 B级（修改），A533 B级1级和A508 2级的RPV钢进行了两种不同的奥氏体化处理和各种热时效处理。在399、425、454和490℃下进行热老化处理，时间分别为168和2000 h。针对每种条件开发了夏比V型缺口冲击韧性与温度的关系曲线，并以韧性-脆性转变温度为基础，比较了各种热处理的效果。很高的奥氏体化热处理会产生非常大的晶粒，这些晶粒在热时效后表现出很高的回火脆化倾向。在许多材料中，沿晶断裂是主要的破坏方式，俄歇分析证实了磷在晶界处的明显偏析。在超级Gleeble中进行的较低温度奥氏体化处理，以模拟埋弧焊件中的原型粗晶粒微结构，产生了预期的晶粒尺寸，其回火脆化的倾向取决于材料以及热时效温度和时间。尽管较低的温度处理降低了回火脆化的可能性，但该结果的确为研究中子辐照和454℃辐照后的热退火导致磷偏析的可能性提供了动力。对其中一种A 302 B级（改性）钢进行了Gleeble处理，在288℃辐照至约0.8 X 10〜（19）n / cm（> 1 MeV），并在454℃进行了168小时的热退火处理。 。在材料的辐照和辐照/退火条件以及接收条件下均进行了夏比冲击试验。结果表明，尽管该材料表现出相对较小的夏比冲击41-J温度变化，但受热影响的区域模拟材料在辐照后退火条件下确实显示出明显的晶间断裂。

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