Modeling and Validation of Irradiation Damage in Ni-based Alloys for Long-Term LWR Applications (US/UK)

长期轻水堆应用镍基合金辐照损伤的建模和验证（美国/英国）

• 批准号: EP/N017854/1
• 负责人: M. Grace Burke
• 金额: $ 52.01万
• 依托单位: University of Manchester
• 依托单位国家: 英国
• 项目类别: Research Grant
• 财政年份: 2016
• 资助国家: 英国
• 起止时间: 2016 至 无数据
• 项目状态: 已结题
• 来源: https://gtr.ukri.org/projects?ref=EP%2FN017854%2F1
• 关键词: Modeling; Validation; Irradiation; Damage; Ni

As pressurised water reactor (PWR) plant operating lives are extended, there is an increased need for predictive modeling tools to describe materials degradation in order to ensure safe, reliable operation as well as plan for component replacements. Some models presently exist, but are limited in their applicability, and are not able to predict degradation of all alloys used in PWR systems. Specifically, it is important to be able to predict any degradation in Ni-based alloys used in nuclear power plants, thus, this program represents an integrated approach to address thermal and irradiation-induced transformations mechanisms of to important Ni-base materials used in PWRs, Alloys 690 and 625. Alloy 690 is widely used in existing PWR plants due to its superior SCC resistance compared to Alloy 600. Alloy 625 is currently used in more limited applications but offers the benefits of both high strength (in the aged condition) and corrosion/SCC resistance, and is being considered for use in future reactor systems. Research has shown that both alloys can undergo phase changes due to thermal or irradiation exposure. In the precipitation-hardened condition, Alloy 625 "softens" during neutron irradiation as the strengthening precipitates decompose and metastable precipitates form. However, the nature and rates of these transformations as a function of exposure conditions are not well understood. Similarly, the effects of these thermal and irradiation-induced microstructural changes on mechanical properties require evaluation. The proposed program combines thermal and irradiation experiments, mechanical testing, and advanced microstructural characterization using state-of-the art analytical techniques, the results of which will be combined with atomistic, micro-and macro-scale modelling that can be used to predict materials performance. Such a capability will also greatly aid in research to develop optimised existing alloys or new alloys for future nuclear power plants.

随着压水反应堆（PWR）电厂运行寿命的延长，越来越需要预测建模工具来描述材料降解，以确保安全可靠的运行以及组件更换计划。目前存在一些模型，但其适用性有限，并且不能预测PWR系统中使用的所有合金的劣化。具体而言，重要的是能够预测核电厂中使用的镍基合金的任何退化，因此，该计划代表了一种综合方法，以解决PWR中使用的重要镍基材料合金690和625的热和辐射诱导转化机制。合金690由于其比合金600更上级的抗应力腐蚀开裂性而广泛用于现有的压水堆核电厂。合金625目前用于更有限的应用，但提供高强度（在老化条件下）和耐腐蚀/SCC的优点，并被考虑用于未来的反应堆系统。研究表明，这两种合金都可以由于热或辐射暴露而发生相变。在沉淀硬化条件下，合金625在中子辐照过程中“软化”，因为强化沉淀物分解并形成亚稳沉淀物。然而，这些转化的性质和速率作为暴露条件的函数还没有很好地理解。同样，这些热和辐射引起的微观结构变化对机械性能的影响需要评估。拟议的计划结合了热和辐照实验，机械测试和先进的微观结构表征，使用最先进的分析技术，其结果将与原子，微观和宏观尺度的建模相结合，可用于预测材料的性能。这种能力也将极大地帮助研究开发优化的现有合金或用于未来核电站的新合金。

项目成果

The application of in situ analytical transmission electron microscopy to the study of preferential intergranular oxidation in Alloy 600.

• DOI: 10.1016/j.ultramic.2016.11.014
• 发表时间: 2017-05
• 期刊: Ultramicroscopy
• 影响因子: 2.2
• 作者: M. G. Burke;G. Bertali;E. Prestat;F. Scenini;S. J. Haigh

Proceedings of the 18th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors

第十八届核电系统材料环境降解国际会议 - 水反应堆

• DOI: 10.1007/978-3-319-67244-1_16
• 发表时间: 2018
• 作者: Teng F

Correlative Microscopy: Elucidating the Mechanisms of SCC in Structural Alloys in PWR Environments

相关显微镜：阐明压水堆环境中结构合金中应力腐蚀开裂的机制

• DOI: 10.1017/s1431927620013665
• 发表时间: 2020
• 期刊: Microscopy and Microanalysis
• 影响因子: 2.8
• 作者: Burke M

The beneficial effect of surface carbon coating on stress corrosion cracking of Type 304 austenitic stainless steels in high temperature water

• DOI: 10.1016/j.scriptamat.2018.08.033
• 发表时间: 2019
• 期刊: Scripta Materialia
• 影响因子: 6
• 作者: K. Mukahiwa;G. Bertali;M. G. Burke;J. Duff;F. Scenini

Transmission Electron Microscopy

透射电子显微镜

• DOI: 10.1007/978-3-319-26651-0_3
• 发表时间: 2016
• 作者: Burke M