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International Journal of Minerals,Metallurgy and Materials 2021年第5期28卷 867-876页

作者： Noora Hytönen Zai-qing Que Pentti Arffman Jari Lydman Pekka Nevasmaa Ulla Ehrnstén Pål Efsing Nuclear reactor Materials VTT Technical Research Centre of Finland Ltd.Kivimiehentie 3P.O.Box 1000FI-02044 VTTFinland Department of Solid Mechanics Royal Institute of Technology(KTH)SE-10044 StockholmSweden

Effects of the weld microstructure and inclusions on brittle fracture initiation are investigated in a thermally aged ferritic high-nickel weld of a reactor pressure vessel head from a decommissioned nuclear power plant.As-welded and reheated regions mainly consist of acicular and polygonal ferrite,respectively.Fractographic examination of Charpy V-notch impact toughness specimens reveals large inclusions(0.5-2.5μm)at the brittle fracture primary initiation sites.High impact energies were measured for the specimens in which brittle fracture was initiated from a small inclusion or an inclusion away from the V-notch.The density,geometry,and chemical composition of the primary initiation inclusions were investigated.A brittle fracture crack initiates as a microcrack either within the multiphase oxide inclusions or from the debonded interfaces between the uncracked inclusions and weld metal matrix.Primary fracture sites can be determined in all the specimens tested in the lower part of the transition curve at and below the 41-J reference impact toughness energy but not above the mentioned value because of the changes in the fracture mechanism and resulting changes in the fracture appearance.

关键词： reactor pressure vessel brittle fracture weld microstructure thermal aging

Investigation on bonding interfaces of an SA508 steel billet manufactured by additive forging

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Journal of Iron and Steel Research(International) 2022年第12期29卷 2016-2023页

作者： Xin-jin Zhang Tian-sheng Wang Zhi-chao Zhu Lin Zhu State Key Laboratory of Metastable Materials Science and Technology Yanshan UniversityQinhuangdao 066004HebeiChina China First Heavy Industries Co. Ltd.Qiqihaer 161042HeilongjiangChina

An experimental steel billet of SA508 reactor pressure vessel material was manufactured by the additive forging method,and microstructure and mechanical properties of the hot-compression bonding interface were systematically investigated.The result indicated that oxidation levels of bonding interfaces were well controlled using vacuum electron beam welding.It was easy to discriminate interfaces from base materials during the optical microstructure observation,since interfaces were characterized by grain or phase boundaries in a straight line.Test results of uniaxial tensile experiments(at 20 and 350°C)and Charpy V-notched impact tests(at 0 and 20°C)showed that fracture behaviour of all those samples appeared at the base material,and bonding interfaces showed advantage of strength and toughness at the forge bonding state.

关键词： reactor pressure vessel Additive forging Interfacial microstructure Mechanical property

OKMC simulation of vacancy-enhanced Cu solute segregation affected by temperature/irradiation in the Fe–Cu system

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Nuclear Science and Techniques 2022年第11期33卷 158-169页

作者： Zi-Qin Shen Jie Gao Sha-Sha Lv Liang Chen Dong-Yue Chen De-Sheng Ai Zheng-Cao Li Key Lab of Advanced Materials(MOE) School of Materials Science and EngineeringTsinghua UniversityBeijing 100084China Nuclear and New Energy Institute Tsinghua UniversityBeijing 100084China Department of Engineering Physics Tsinghua UniversityBeijing 100084China Key Laboratory of Beam Technology Ministry of EducationCollege of Nuclear Science and TechnologyBeijing Normal UniversityBeijing 100875China School of Materials Science and Engineering Shanghai Jiao Tong UniversityShanghai 200240China Department of Nuclear Engineering and Management The University of TokyoTokyo 113-8656Japan

The effects of annealing and irradiation on the evolution of Cu clusters in a-Fe are investigated using object kinetic Monte Carlo simulations.In our model,vacancies act as carriers for chemical species via thermally activated diffusion jumps,thus playing an important role in solute diffusion.At the end of the Cu cluster evolution,the simulations of the average radius and number density of the clusters are consistent with the experimental data,which indicates that the proposed simulation model is applicable and effective.For the simulation of the annealing process,it is found that the evolution of the cluster size roughly follows the 1/2 time power law with the increase in radius during the growth phase and the 1/3 time power law during the coarsening phase.In addition,the main difference between neutron and ion irradiation is the growth and evolution process of the copper-vacancy clusters.The aggregation of vacancy clusters under ion irradiation suppresses the migration and coarsening of the clusters,which ultimately leads to a smaller average radius of the copper clusters.Our proposed simulation model can supplement experimental analyses and provide a detailed evolution mechanism of vacancy-enhanced precipitation,thereby providing a foundation for other elemental precipitation research.

关键词： Object kinetic Monte Carlo Irradiation effect Solute segregation reactor pressure vessel

A Peridynamic Approach for the Evaluation of Metal Ablation under High Temperature

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Computer Modeling in Engineering & Sciences 2023年第3期134卷 1997-2019页

作者： Hui Li Liping Zhang Yixiong Zhang Xiaolong Fu Xuejiao Shao Juan Du Science and Technology on reactor System Design Technology Laboratory Nuclear Power Institute of ChinaChengdu610200China

In this paper,the evaluations of metal ablation processes under high temperature,i.e.,the Al plate ablated by a laser and a heat carrier and the reactor pressure vessel ablated by a core melt,are studied by a novel peridynamic method.Above all,the peridynamic formulation for the heat conduction problem is obtained by Taylor’s expansion technique.Then,a simple and efficient moving boundary model in the peridynamic framework is proposed to handle the variable geometries,in which the ablated states of material points are described by an additional scalar field.Next,due to the automatic non-interpenetration properties of peridynamic method,a contact algorithm is established to determine the contact relationship between the ablated system and the additional heat carrier.In addition,the corresponding computational procedure is listed in detail.Finally,several numerical examples are carried out and the results verify the validity and accuracy of the present method.

关键词： Peridynamics metal ablation moving boundary model contact algorithm reactor pressure vessel

Improvement in irradiation resistance of FeCu alloy by pre-deformation through introduction of dense point defect sinks

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Rare Metals 2021年第4期40卷 885-896页

作者： Hao Wu Qiu-Lin Li Ben Xu Hai-Long Liu Guo-Gang Shu Wei Liu Graduate School at Shenzhen Tsinghua UniversityShenzhen 518055China School of Materials Science and Engineering Tsinghua UniversityBeijing 100084China State Key Laboratory of Nuclear Power Safety Monitoring Technology and Equipment China Nuclear Power Engineering Co.LtdShenzhen 518172China

The irradiation resistance of pre-deformed FeCu alloy was studied using a 3 MeV Fe ion irradiation experiment at room temperature in comparison with that of the as-received sample.Nanoindentation and atom probe tomography(APT) were used to characterize the mechanical properties and solute distribution.The stress-strain curve obtained by nanoindentation shows that the yield strength(σ0.2) of the pre-deformed sample is unexpectedly reduced with an increase in the irradiation dose to five displacements per atom(dpa).We suggest that it results both from the decrease in the dislocation density and the suppression of defects during irradiation.APT shows that the nucleation of the Cu cluster is suppressed;however,its growth is promoted in the pre-deformed sample,resulting in the formation of sparse and coarse clusters at 1 dpa irradiation.These coarse Cu clusters were then unexpectedly refined to finer grains with an increase in the irradiation dose to 5 dpa.Theoretically,the improvement in the resistance to irradiation in the pre-deformed sample is attributed to the dense point-defect sinks,that is,the dislocations and grain boundaries introduced by pre-deformation.In addition,the contributions of the dislocations and grain boundaries to the sink strength are estimated for both the as-received and pre-deformed samples.The results indicate that dislocations,rather than grain boundaries,play a major role after deformation.

关键词： reactor pressure vessel Irradiation damage Pre-deformation Dislocation density Cu-rich clusters

Influence evaluation of loading conditions during pressurized thermal shock transients based on thermal-hydraulics and structural analyses

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Frontiers of Mechanical Engineering 2018年第4期13卷 563-570页

作者： Jinya KATSUYAMA Shumpei UNO Tadashi WATANABE Yinsheng LI Japan Atomic Energy Agency Ibaraki 319-1195 Japan Mizuho Information & Research Institute Inc. Tokyo 101-8443 Japan Fukui University Fukui 914-0055 Japan

The thermal hydraulic （TH） behavior of coo- lant water is a key factor in the structural integrity assessments on reactor pressure vessels （RPVs） of pressurized water reactors （PWRs） under pressurized thermal shock （PTS） events, because the TH behavior may affect the loading conditions in the assessment. From the viewpoint of TH behavior, configuration of plant equipment and their dimensions, and operator action time considerably influence various parameters, such as the temperature and flow rate of coolant water and inner pressure. In this study, to investigate the influence of the operator action time on TH behavior during a PTS event, we developed an analysis model for a typical Japanese PWR plant, including the RPV and the main components of both primary and secondary systems, and performed TH analyses by using a system analysis code called RELAP5. We applied two different operator action times based on the Japanese and the United States （US） rules： Operators may act after 10 min （Japanese rules） and 30 min （the US rules） after the occurrence of PTS events. Based on the results of TH analysis with different operator action times, we also performed structural analyses for evaluating thermal-stress distributions in the RPV during PTS events as loading conditions in the structural integrity assessment. From the analysis results, it was clarified that differences in operator action times significantly affect TH behavior and loading conditions, as the Japanese rule may lead to lower stresses than that under the US rule because an earlier operator action caused lower pressure in the RPV.

关键词： structural integrity reactor pressure vessel pressurized thermal shock thermal hydraulic analysis pressurized water reactor weld residual stress

Microscopic damage mechanism of SA508 Gr3 steel in ultra-high temperature creep

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Journal of Iron and Steel Research(International) 2018年第4期25卷 453-459页

作者： Zhi-gang Xie Yan-ming He Jian-guo Yang Xiang-qing Li Chuan-yang Lu Zeng-liang Gao Institute of Process Equipment and Control Engineering Zhejiang University of Technology Hangzhou 310014 Zhejiang China Mechanical and Electrical Department Shantou Polytechnic College Shantou 515078 Guangdong China

The lower head of reactor pressure vessel （RPV） will endure a great temperature gradient above the phase transition temperature, and the creep and fracture will be the primary failure mode for the RPV material in such a situation. The interrupted creep tests were performed on a typical RPV material, SA508 Gr3 steel, at 800 ℃. The microstructure of different creep stages was examined by scanning electron microscopy and transmission electron microscopy. The results showed that the microscopic damage is mainly induced by creep cavities and coarse second-phase particles. Furthermore, the volume fractions of creep cavities and coarse second-phase particles show a linear relationship with the extended creep time. The second-phase particles are determined to be MoC in the second creep stage and Mo2C in the third creep stage, according to the results of selected-area electron diffraction pattern. Combined with energy-dispersive spectrum analysis, the segregation of precipitates caused by the migration of atoms is finally unveiled, which leads to the coarsening of the particles.

关键词： reactor pressure vessel SA508 Gr3 steel In-vessel retention. Second-phase particle Damage mechanism Ultra-high temperature creep

A modified theta projection model for creep behavior of RPV steel 16MND5

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Journal of Materials Science & Technology 2020年第12期47卷 231-242页

作者： Peng Yu Weimin Ma Division of Nuclear Power Safety Royal Institute of Technology(KTH)Stockholm 10691Sweden

During a hypothetical severe accident of light water reactors,the reactor pressure vessel(RPV) could fail due to its creep under the influence of high-temperature corium.Hence,modelling of creep behavior of the RPV is paramount to reactor safety analysis since it predicts the transition point of accident progression from in-vessel to ex-vessel phase.In the present study we proposed a new creep model for the classical French RPV steel 16 MND5,which is adapted from the "theta-projection model" and contains all three stages of a creep process.Creep curves are expressed as a function of time with five model parameters θ_i(i=1-4 and m).A model parameter dataset was constructed by fitting experimental creep curves into this function.To correlate the creep curves for different temperatures and stress loads,we directly interpolate the model’s parameters θ_i(i=1-4 and m) from this dataset,in contrast to the conventional "theta-projection model" which employs an extra single correlation for each θ_i(i=1-4 andm),to better accommodate all experimental curves over the wide ranges of temperature and stress loads.We also put a constraint on the trend of the creep strain that it would monotonically increase with temperature and stress load.A good agreement was achieved between each experimental creep curve and corresponding model’s prediction.The widely used time-hardening and strain-hardening models were performing reasonably well in the new method.

关键词： 16MND5 steel Creep modelling Tertiary stage reactor pressure vessel Theta projection model