{"title":"Creep deformation and rupture behavior of 9Cr-ODS steel cladding tube at high temperatures from 700°C to 1000°C","authors":"Yuya Imagawa, Ryuta Hashidate, Takeshi Miyazawa, Takashi Onizawa, Satoshi Ohtsuka, Yasuhide Yano, Takashi Tanno, Takeji Kaito, Masato Ohnuma, Masatoshi Mitsuhara, Takeshi Toyama","doi":"10.1080/00223131.2023.2269178","DOIUrl":null,"url":null,"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.","PeriodicalId":16526,"journal":{"name":"Journal of Nuclear Science and Technology","volume":"162 1","pages":"0"},"PeriodicalIF":1.5000,"publicationDate":"2023-10-19","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":"0","resultStr":null,"platform":"Semanticscholar","paperid":null,"PeriodicalName":"Journal of Nuclear Science and Technology","FirstCategoryId":"1085","ListUrlMain":"https://doi.org/10.1080/00223131.2023.2269178","RegionNum":4,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":null,"EPubDate":"","PubModel":"","JCR":"Q2","JCRName":"NUCLEAR SCIENCE & TECHNOLOGY","Score":null,"Total":0}
引用次数: 0
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.
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.