• Title/Summary/Keyword: Accident-tolerant fuel

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Thermal-Hydraulic Research Review and Cooperation Outcome for Light Water Reactor Fuel (경수로핵연료 열수력 연구개발 분석 및 연산학 협력 성과)

  • In, Wang Kee;Shin, Chang Hwan;Lee, Chi Young;Lee, Chan;Chun, Tae Hyun;Oh, Dong Seok
    • Transactions of the Korean Society of Mechanical Engineers B
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    • v.40 no.12
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    • pp.815-824
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    • 2016
  • The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermal-hydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermal-hydraulic technology and the commercialization.

EBSD studies on microstructure and crystallographic orientation of UO2-Mo composite fuels

  • Tummalapalli, Murali Krishna;Szpunar, Jerzy A.;Prasad, Anil;Bichler, Lukas
    • Nuclear Engineering and Technology
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    • v.53 no.12
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    • pp.4052-4059
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    • 2021
  • The microstructure of the fuel pellet plays an essential role in fission gas buildup and release and is critical for the safe and continued operation of nuclear power stations. Structural analysis of uranium dioxide (UO2)-molybdenum (Mo) composite fuel pellets prepared at a range of sintering temperatures from 1300 to 1800 ℃ was performed. Mo micro and nanoparticles were used in making the composite pellets. A systematic investigation into the influence of processing parameters during Spark Plasma Sintering (SPS) of the pellets on the microstructure, texture, grain size, and grain boundary characters of UO2-Mo is presented. UO2-Mo composite show significant differences in the fraction of general boundaries and also special/coincident site lattice (CSL) boundaries. EBSD orientation maps demonstrated that <111> texturing was observed in the pellets fabricated at 1500 ℃. The experimental investigations suggest that UO2-Mo composite pellets have favorable microstructural features compared to the UO2 pellet.

Neutronic design and evaluation of the solid microencapsulated fuel in LWR

  • Deng, Qianliang;Li, Songyang;Wang, Dingqu;Liu, Zhihong;Xie, Fei;Zhao, Jing;Liang, Jingang;Jiang, Yueyuan
    • Nuclear Engineering and Technology
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    • v.54 no.8
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    • pp.3095-3105
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    • 2022
  • Solid Microencapsulated Fuel (SMF) is a type of solid fuel rod design that disperses TRISO coated fuel particles directly into a kind of matrix. SMF is expected to provide improved performance because of the elimination of cladding tube and associated failure mechanisms. This study focused on the neutronics and some of the fuel cycle characteristics of SMF by using OpenMC. Two kinds of SMFs have been designed and evaluated - fuel particles dispersed into a silicon carbide matrix and fuel particles dispersed into a zirconium matrix. A 7×7 fuel assembly with increased rod diameter transformed from the standard NHR200-II 9×9 array was also introduced to increase the heavy metal inventory. A preliminary study of two kinds of burnable poisons (Erbia & Gadolinia) in two forms (BISO and QUADRISO particles) was also included. This study found that SMF requires about 12% enriched UN TRISO particles to match the cycle length of standard fuel when loaded in NHR200-II, which is about 7% for SMF with increased rod diameter. Feedback coefficients are less negative through the life of SMF than the reference. And it is estimated that the average center temperature of fuel kernel at fuel rod centerline is about 60 K below that of reference in this paper.

Design and evaluation of an innovative LWR fuel combined dual-cooled annular geometry and SiC cladding materials

  • Deng, Yangbin;Liu, Minghao;Qiu, Bowen;Yin, Yuan;Gong, Xing;Huang, Xi;Pang, Bo;Li, Yongchun
    • Nuclear Engineering and Technology
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    • v.53 no.1
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    • pp.178-187
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    • 2021
  • Dual-cooled annular fuel allows a significant increase in power density while maintaining or improving safety margins. However, the dual-cooled design brings much higher Zircaloy charge in reactor core, which could cause a great threaten of hydrogen explosion during severe accidents. Hence, an innovative fuel combined dual-cooled annular geometry and SiC cladding was proposed for the first time in this study. Capabilities of fuel design and behavior simulation were developed for this new fuel by the upgrade of FROBA-ANNULAR code. Considering characteristics of both SiC cladding and dual-cooled annular geometry, the basic fuel design was proposed and preliminary proved to be feasible. After that, a design optimization study was conducted, and the optimal values of as-fabricated plenum pressure and gas gap sizes were obtained. Finally, the performance simulation of the new fuel was carried out with the full consideration of realistic operation conditions. Results indicate that in addition to possessing advantages of both dual-cooled annular fuel and accident tolerant cladding at the same time, this innovative fuel could overcome the brittle failure issue of SiC induced by pellet-cladding interaction.

A review on thermohydraulic and mechanical-physical properties of SiC, FeCrAl and Ti3SiC2 for ATF cladding

  • Qiu, Bowen;Wang, Jun;Deng, Yangbin;Wang, Mingjun;Wu, Yingwei;Qiu, S.Z.
    • Nuclear Engineering and Technology
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    • v.52 no.1
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    • pp.1-13
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    • 2020
  • At present, the Department of Energy (DOE) in Unite State are directing the efforts of developing accident tolerant fuel (ATF) technology. As the first barrier of nuclear fuel system, the material selection of fuel rod cladding for ATFs is a basic but very significant issue for the development of this concept. The advanced cladding is attractive for providing much stronger oxidation resistance and better in-pile behavior under sever accident conditions (such as SBO, LOCA) for giving more coping time and, of course, at least an equivalent performance under normal condition. In recent years, many researches on in-plie or out-pile physical properties of some suggested cladding materials have been conducted to solve this material selection problem. Base on published literatures, this paper introduced relevant research backgrounds, objectives, research institutions and their progresses on several main potential claddings include triplex SiC, FeCrAl and MAX phase material Ti3SiC2. The physical properties of these claddings for their application in ATF area are also reviewed in thermohydraulic and mechanical view for better understanding and simulating the behaviors of these new claddings. While most of important data are available from publications, there are still many relevant properties are lacking for the evaluations.

Mechanical analysis of surface-coated zircaloy cladding

  • Lee, Youho;Lee, Jeong Ik;NO, Hee Cheon
    • Nuclear Engineering and Technology
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    • v.49 no.5
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    • pp.1031-1043
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    • 2017
  • A structural model for stress distributions of coated Zircaloy subjected to realistic incore pressure difference, thermal expansion, irradiation-induced axial growth, and creep has been developed in this study. In normal operation, the structural integrity of coating layers is anticipated to be significantly challenged with increasing burnup. Strain mismatch between the zircaloy and the coated layer, due to their different irradiation-induced axial growth, and creep deformation are found to be the most dominant causes of stress. This study suggests that the compatibility of the high temperature irradiation-induced strains (axial growth and creep) between zircaloy and the coating layer and the capability to undergo plastic strain should be taken as key metrics, along with the traditional focus on chemical protectiveness.

Improved evaluation of ring tensile test ductility applied to neutron irradiated 42XNM tubes in the temperature range of (500-1100)℃

  • Gurovich, B.A.;Frolov, A.S.;Fedotov, I.V.
    • Nuclear Engineering and Technology
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    • v.52 no.6
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    • pp.1213-1221
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    • 2020
  • Chromium-nickel alloy 42XNM (XHM-1, Bochvalloy) is considered as a promising material for future generations of nuclear reactors, primarily as a material for the fuel elements shells in the development of accident tolerant fuel. However, as with most nickel-based alloys, 42ХNМ is characterized by a sharp decrease in plastic properties in the temperature range of (500-900)℃. This effect is enhanced by neutron irradiation. Preliminary tests of ring samples of 42XNM alloy (after irradiation as a part of the VVER-1000 control system) in the temperature range of ductility failure showed that the standard technique for processing tensile diagrams does not allow to evaluate the plastic properties correctly at low strains. Therefore, in this work, the alternative method for testing ring samples from materials with low plastic characteristics was developed. It was shown that the minimum value of the permanent strain of the irradiated 42XNM alloy in the temperature range of (500-1100)℃, determined by the alternative method, was ~1.6% at 750 ℃.

Resistance, electron- and laser-beam welding of zirconium alloys for nuclear applications: A review

  • Slobodyan, Mikhail
    • Nuclear Engineering and Technology
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    • v.53 no.4
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    • pp.1049-1078
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    • 2021
  • The review summarizes the published data on the widely applied electron-beam, laser-beam, as well as resistance upset, projection, and spot welding of zirconium alloys for nuclear applications. It provides the results of their analysis to identify common patterns in this area. Great attention has been paid to the quality requirements, the edge preparation, up-to-date equipment, process parameters, as well as post-weld treatment and processing. Also, quality control and weld repair methods have been mentioned. Finally, conclusions have been drawn about a significant gap between the capabilities of advanced welding equipment to control the microstructure and, accordingly, the properties of welded joints of the zirconium alloys and existing algorithms that enable to realize them in the nuclear industry. Considering the ever-increasing demands on the high-burnup accident tolerant nuclear fuel assemblies, great efforts should be focused on the improving the welding procedures by implementing predefined heat input cycles. However, a lot of research is required, since the number of possible combinations of the zirconium alloys, designs and dimensions of the joints dramatically exceeds the quantity of published results on the effect of the welding parameters on the properties of the welds.

Investigation on effect of surface properties on droplet impact cooling of cladding surfaces

  • Wang, Zefeng;Qu, Wenhai;Xiong, Jinbiao;Zhong, Mingjun;Yang, Yanhua
    • Nuclear Engineering and Technology
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    • v.52 no.3
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    • pp.508-519
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    • 2020
  • During transients or accidents, the reactor core is uncovered, and droplets entrained above the quench front collides with the uncovered fuel rod surface. Droplet impact cooling can reduce the peak cladding temperature. Besides zirconium-based cladding, versatile accidental tolerant fuel (ATF) claddings, including FeCrAl, have been proposed to increase the accident coping time. In order to investigate the effect of surface properties on droplet impact cooling of cladding surfaces, the droplet impact phenomena are photographed on the FeCrAl and zircaloy-4 (Zr-4) surfaces under different conditions. On the oxidized FeCrAl surface, the Leidenfrost phenomenon is not observed even when the surface temperature is as high as 550 ℃ with We > 30. Comparison of the impact behaviors observed on different materials shows that nucleate and transition boiling is more intensive on surfaces with larger thermal conductivity. The Leidenfrost point temperature (LPT) decreases with the solid thermal effusivity (${\sqrt{k{\rho}C_p}}$). However, the CHF temperature is relatively insensitive to the surface oxidation and Weber number. Droplet spreading diameter is analyzed quantitatively in the film boiling stage. Based on the energy balance a correlation is proposed for droplet maximum spreading factor. A mechanistic model is also developed for the LPT based on homogeneous nucleation theory.