首页> 外文会议>International congress on advances in nuclear power plants >ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO U.S. MATERIAL TEST REACTORS
【24h】

ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO U.S. MATERIAL TEST REACTORS

机译:裂纹扩展检测技术适应美国材料测试反应器

获取原文

摘要

Idaho National Laboratory and the Massachusetts Institute of Technology Nuclear Reactor Laboratory researchers have recently completed a project to develop a test rig capable of evaluating the potential for irradiation-assisted stress corrosion cracking in the high-flux conditions possible in U.S. material test reactors. Adapting techniques pioneered at the Halden Boiling Water Reactor, the project included designing and testing the loader mechanism, testing individual test rig components and electronics support equipment, and autoclave testing of the rig design. Technical challenges involved developing robust connections to the specimen for the applied current and voltage measurements, selecting appropriate ceramic insulating materials that can endure light-water reactor environments, coping with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, primarily associated with fast-neutron damage that changes specimen resistance without additional crack growth. The project culminated with an in-pile demonstration at the Massachusetts Institute of Technology Research Reactor. The test rig and associated support equipment were used to apply loads to a representative compact tensile specimen and measure crack growth using the direct current potential drop method. Although the test period was limited to approximately 70 days and the accumulated neutron dose was relatively small, successful operation of the test rig was demonstrated. The specimen was cycled more than 8,000 times (more than typical for a long-term irradiation-assisted stress corrosion cracking test), which was sufficient to propagate a crack of over 2 mm. Post-irradiation measurements of the specimen confirmed a crack length equivalent to that inferred by the direct current potential drop signals. Measurements also demonstrated the necessary infrastructure for handling the irradiated sample rig and satisfactory performance of a low-cost system for crack measurement in the hot cell.
机译:爱达荷州国家实验室和麻省理工学院核反应堆实验室的研究人员最近完成了一个开发试验台的项目,该试验台能够评估在美国材料试验反应堆可能出现的高通量条件下辐照辅助应力腐蚀开裂的可能性。该项目采用了Halden沸水反应堆的先驱技术,包括设计和测试装载机机构,测试各个测试平台的组件和电子支持设备,以及对设计的高压釜进行测试。技术挑战包括开发与样品的牢固连接以进行施加的电流和电压测量,选择可承受轻水反应堆环境的合适陶瓷绝缘材料,以全功率应对反应堆堆芯的高电磁噪声环境以及适应材料性能试样的变化,主要与快中子损坏有关,后者会改变试样的电阻而不会产生额外的裂纹。该项目最终在麻省理工学院反应堆进行了堆内演示。测试设备和相关的支撑设备用于将载荷施加到代表性的紧凑型拉伸试样上,并使用直流电势下降法测量裂纹的扩展。尽管测试周期被限制为大约70天,并且累积的中子剂量相对较小,但是却证明了该测试装置的成功运行。样品循环了8,000次以上(比长期辐照辅助应力腐蚀开裂测试的典型次数还要多),足以传播超过2毫米的裂纹。样品的辐照后测量证实了裂纹长度等于直流电势下降信号所推断的裂纹长度。测量还证明了处理辐照样品台的必要基础设施以及低成本系统在热室中进行裂纹测量的令人满意的性能。

著录项

相似文献

  • 外文文献
  • 中文文献
  • 专利
获取原文

客服邮箱:kefu@zhangqiaokeyan.com

京公网安备:11010802029741号 ICP备案号:京ICP备15016152号-6 六维联合信息科技 (北京) 有限公司©版权所有
  • 客服微信

  • 服务号