Creep deformation and rupture behavior of 9Cr-ODS steel cladding tube at high temperatures from 700°C to 1000°C

IF 1.5 4区 工程技术 Q2 NUCLEAR SCIENCE & TECHNOLOGY Journal of Nuclear Science and Technology Pub Date : 2023-10-19 DOI:10.1080/00223131.2023.2269178
Yuya Imagawa, Ryuta Hashidate, Takeshi Miyazawa, Takashi Onizawa, Satoshi Ohtsuka, Yasuhide Yano, Takashi Tanno, Takeji Kaito, Masato Ohnuma, Masatoshi Mitsuhara, Takeshi Toyama
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Abstract

ABSTRACTThe Japan Atomic Energy Agency has been developing 9Cr-oxide dispersion strengthened (ODS) steel as a fuel cladding material for sodium-cooled fast reactors (SFRs). Previous studies have formulated the creep rupture equation for 650°C–850°C. However, little data have been obtained above 850°C, and no equation has been formulated. This study conducted creep tests to evaluate creep strength at 700°C–1000°C. Two creep test methods, the internal pressure and ring creep tests under development, were used, and the validation of the ring creep test method was conducted. The results showed that 9Cr-ODS steel undergoes almost no strength change due to the matrix’s phase transformation, and a single equation can express a creep rupture strength from 700°C to 1000°C. In validating the ring creep test method, analysis clarified the effect of stress concentration on the specimen. Plastic deformation occurs at high initial stress and may lead to early rupture. The results will be essential for future creep testing and evaluation of neutron-irradiated 9Cr-ODS steel.KEYWORDS: Oxide dispersion strengthened steelfuel cladding tube,creep strengthcreep straininternal creep testring creep testDisclaimerAs a service to authors and researchers we are providing this version of an accepted manuscript (AM). Copyediting, typesetting, and review of the resulting proofs will be undertaken on this manuscript before final publication of the Version of Record (VoR). During production and pre-press, errors may be discovered which could affect the content, and all legal disclaimers that apply to the journal relate to these versions also. AcknowledgmentsThe authors would like to express their sincere gratitude to Dr. Tomoyuki Uwaba for his valuable guidance on finite element simulation.Additional informationFundingMEXT Innovative Nuclear Research and Development Program Grant Number JPMXD0219214482, Ministry of Education Culture, Sports, Science, and Technology, supported this work.
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9Cr-ODS钢包层管在700 ~ 1000℃高温下的蠕变变形和断裂行为
摘要日本原子能机构一直致力于开发9cr -氧化物弥散强化(ODS)钢作为钠冷快堆(SFRs)燃料包壳材料。已有研究建立了650℃- 850℃蠕变断裂方程。然而,850℃以上的数据很少,也没有公式。本研究进行了蠕变试验,以评估700°C - 1000°C的蠕变强度。采用正在开发的内压蠕变试验和环蠕变试验两种蠕变试验方法,对环蠕变试验方法进行了验证。结果表明:9Cr-ODS钢的强度几乎不受基体相变的影响,700 ~ 1000℃范围内的蠕变断裂强度可以用单一公式表示。在验证环蠕变试验方法时,分析了应力集中对试件的影响。塑性变形发生在高初始应力下,可能导致早期断裂。研究结果对今后中子辐照9Cr-ODS钢的蠕变试验和评价具有重要意义。关键词:氧化物弥散强化钢燃料包壳管,蠕变强度,蠕变应变,内部蠕变试验管柱蠕变试验免责声明作为对作者和研究人员的服务,我们提供此版本的接受稿件(AM)。在最终出版版本记录(VoR)之前,将对该手稿进行编辑、排版和审查。在制作和印前,可能会发现可能影响内容的错误,所有适用于期刊的法律免责声明也与这些版本有关。作者衷心感谢Tomoyuki Uwaba博士在有限元模拟方面的宝贵指导。文部省创新核研究与发展计划资助号:JPMXD0219214482,教育、文化、体育、科学和技术部支持这项工作。
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来源期刊
Journal of Nuclear Science and Technology
Journal of Nuclear Science and Technology 工程技术-核科学技术
CiteScore
2.40
自引率
16.70%
发文量
116
审稿时长
2.3 months
期刊介绍: The Journal of Nuclear Science and Technology (JNST) publishes internationally peer-reviewed papers that contribute to the exchange of research, ideas and developments in the field of nuclear science and technology, to contribute peaceful and sustainable development of the World. JNST ’s broad scope covers a wide range of topics within its subject category, including but are not limited to: General Issues related to Nuclear Power Utilization: Philosophy and Ethics, Justice and Policy, International Relation, Economical and Sociological Aspects, Environmental Aspects, Education, Documentation and Database, Nuclear Non-Proliferation, Safeguard Radiation, Accelerator and Beam Technologies: Nuclear Physics, Nuclear Reaction for Engineering, Nuclear Data Measurement and Evaluation, Integral Verification/Validation and Benchmark on Nuclear Data, Radiation Behaviors and Shielding, Radiation Physics, Radiation Detection and Measurement, Accelerator and Beam Technology, Synchrotron Radiation, Medical Reactor and Accelerator, Neutron Source, Neutron Technology Nuclear Reactor Physics: Reactor Physics Experiments, Reactor Neutronics Design and Evaluation, Reactor Analysis, Neutron Transport Calculation, Reactor Dynamics Experiment, Nuclear Criticality Safety, Fuel Burnup and Nuclear Transmutation, Reactor Instrumentation and Control, Human-Machine System: Reactor Instrumentation and Control System, Human Factor, Control Room and Operator Interface Design, Remote Control, Robotics, Image Processing Thermal Hydraulics: Thermal Hydraulic Experiment and Analysis, Thermal Hydraulic Design, Thermal Hydraulics of Single/Two/Multi Phase Flow, Interactive Phenomena with Fluid, Measurement Technology...etc.
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