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Jernkvist LO. A review of analytical criteria for fission gas induced fragmentation of oxide fuel in accident conditions. Progress in Nuclear Energy 2020. [DOI: 10.1016/j.pnucene.2019.103188] [Citation(s) in RCA: 9] [Impact Index Per Article: 2.3] [Reference Citation Analysis] [What about the content of this article? (0)] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 11/30/2022]
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Abstract
In reactor accidents that involve rapid overheating of oxide fuel, overpressurization of gas-filled bubbles and pores may lead to rupture of these cavities, fine fragmentation of the fuel material, and burst-type release of the cavity gas. Analytical rupture criteria for various types of cavities exist, but application of these criteria requires that microstructural characteristics of the fuel, such as cavity size, shape and number density, are known together with the gas content of the cavities. In this paper, we integrate rupture criteria for two kinds of cavities with models that calculate the aforementioned parameters in UO2 LWR fuel for a given operating history. The models are intended for implementation in engineering type computer programs for thermal-mechanical analyses of LWR fuel rods. Here, they have been implemented in the FRAPCON and FRAPTRAN programs and validated against experiments that simulate LOCA and RIA conditions. The capabilities and shortcomings of the proposed models are discussed in light of selected results from this validation. Calculated results suggest that the extent of fuel fragmentation and transient fission gas release depends strongly on the pre-accident fuel microstructure and fission gas distribution, but also on rapid changes in the external pressure exerted on the fuel pellets during the accident.
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Itakura M, Nakamura H, Kitagaki T, Hoshino T, Machida M. First-principles calculation of mechanical properties of simulated debris ZrxU1−xO2. J NUCL SCI TECHNOL 2019. [DOI: 10.1080/00223131.2019.1604271] [Citation(s) in RCA: 2] [Impact Index Per Article: 0.4] [Reference Citation Analysis] [What about the content of this article? (0)] [Affiliation(s)] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 10/27/2022]
Affiliation(s)
- Mitsuhiro Itakura
- Center for Computational Science & e-Systems, Japan Atomic Energy Agency, Kashiwa, Japan
| | - Hiroki Nakamura
- Center for Computational Science & e-Systems, Japan Atomic Energy Agency, Kashiwa, Japan
| | - Toru Kitagaki
- Nuclear Fuel Cycle Engineering Laboratories, Japan Atomic Energy Agency, Ibaraki, Japan
| | - Takanori Hoshino
- Nuclear Fuel Cycle Engineering Laboratories, Japan Atomic Energy Agency, Ibaraki, Japan
| | - Masahiko Machida
- Center for Computational Science & e-Systems, Japan Atomic Energy Agency, Kashiwa, Japan
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Kitagaki T, Hoshino T, Yano K, Okamura N, Ohara H, Fukasawa T, Koizumi K. Mechanical Properties of Cubic (U,Zr)O2. Journal of Nuclear Engineering and Radiation Science 2018. [DOI: 10.1115/1.4039847] [Citation(s) in RCA: 7] [Impact Index Per Article: 1.2] [Reference Citation Analysis] [What about the content of this article? (0)] [Affiliation(s)] [Abstract] [Track Full Text] [Journal Information] [Subscribe] [Scholar Register] [Indexed: 11/08/2022]
Abstract
Evaluation of fuel debris properties in the Fukushima Daiichi nuclear power plant (1F) is required to develop fuel debris removal tools. In the removal of debris resulting from the Three Mile Island unit 2 (TMI-2) accident, a core-boring system played an important role. Considering the working principle of core boring, hardness, elastic modulus, and fracture toughness were found to be important fuel debris properties that profoundly influenced the performance of the boring machine. It is speculated that uranium and zirconium oxide solid solution (U,Zr)O2 is one of the major materials in the fuel debris from 1F. In addition, the Zr content of the fuel debris from 1F is expected to be higher than that of the debris from TMI-2 because the 1F reactors were boiling-water reactors. In this research, the mechanical properties of cubic (U,Zr)O2 samples containing 10%–65% ZrO2 are evaluated. The hardness, elastic modulus, and fracture toughness are measured by the Vickers test, ultrasonic pulse echo method, and indentation fracture method, respectively. In the case of (U,Zr)O2 samples containing less than 50% ZrO2, Vickers hardness and fracture toughness increased, and the elastic modulus decreased slightly with increasing ZrO2 content. Moreover, all of those values of the (U,Zr)O2 samples containing 65% ZrO2 increased slightly compared to (U,Zr)O2 samples containing 55% ZrO2. ZrO2 content affects fracture toughness significantly in the case of samples containing less than 10% ZrO2. Higher Zr content (exceeding 50%) has little effect on the mechanical properties.
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Affiliation(s)
- Toru Kitagaki
- Japan Atomic Energy Agency, Nuclear Fuel Cycle Engineering Laboratories, 4-33 Muramatsu, Tokai-mura 319-1194, Ibaraki, Japan e-mail:
| | - Takanori Hoshino
- Japan Atomic Energy Agency, Nuclear Fuel Cycle Engineering Laboratories, 4-33 Muramatsu, Tokai-mura 319-1194, Ibaraki, Japan e-mail:
| | - Kimihiko Yano
- Japan Atomic Energy Agency, Nuclear Fuel Cycle Engineering Laboratories, 4-33 Muramatsu, Tokai-mura 319-1194, Ibaraki, Japan e-mail:
| | - Nobuo Okamura
- Japan Atomic Energy Agency, Nuclear Fuel Cycle Engineering Laboratories, 4-33 Muramatsu, Tokai-mura 319-1194, Ibaraki, Japan e-mail:
| | - Hiroshi Ohara
- Nippon Nuclear Fuel Development Co., LTD., 2163 Narita-cho, Oarai-Machi 311-1313, Ibaraki, Japan e-mail:
| | - Tetsuo Fukasawa
- Hitachi-GE Nuclear Energy, LTD., 3-1-1 Saiwai-cho, Hitachi-Shi 317-0073, Ibaraki, Japan e-mail:
| | - Kenji Koizumi
- Japan Atomic Energy Agency, Nuclear Fuel Cycle Engineering Laboratories, 4-33 Muramatsu, Tokai-mura 319-1194, Ibaraki, Japan e-mail:
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Yao T, Xin G, Scott SM, Gong B, Lian J. Thermally-Conductive and Mechanically-Robust Graphene Nanoplatelet Reinforced UO 2 Composite Nuclear Fuels. Sci Rep 2018; 8:2987. [PMID: 29445176 PMCID: PMC5812999 DOI: 10.1038/s41598-018-21034-4] [Citation(s) in RCA: 16] [Impact Index Per Article: 2.7] [Reference Citation Analysis] [What about the content of this article? (0)] [Affiliation(s)] [Abstract] [Track Full Text] [Download PDF] [Figures] [Journal Information] [Subscribe] [Scholar Register] [Received: 07/14/2017] [Accepted: 01/29/2018] [Indexed: 11/16/2022] Open
Abstract
Low thermal transport behavior along the radial direction of nuclear fuel pellets and pellet-cladding mechanical interaction significantly impact fuel performance and the safety of current nuclear energy systems. Here we report a new strategy of advanced fuel design in which highly thermally-conductive and mechanically-robust graphene nanoplatelets are incorporated into UO2 fuel matrix to improve fuel thermal-mechanical properties. The 2D geometry of the graphene nanoplatelets enables a unique lamellar structure upon fuel consolidation by spark plasma sintering. The thermal conductivity along the radial direction of the sintered fuel pellets at room temperature reaches 12.7 and 19.1 wm−1K−1 at 1 wt.% and 5 wt.% loadings of the graphene nanoplatelets, respectively, representing at least 74% and 162% enhancements as compared to pure UO2 fuel pellets. Indentation testing suggests great capability of the 2D graphene nanoplatelets to deflect and pin crack propagation, drastically improving the crack propagation resistance of fuel matrix. The estimated indentation fracture toughness reaches 3.5 MPa·m1/2 by 1 wt.% loading of graphene nano-platelets, representing a 150% improvement over 1.4 MPa·m1/2 for pure UO2 fuel pellets. Isothermal annealing of the composite fuel indicates that the graphene nano-platelet is able to retain its structure and properties against reaction with UO2 matrix up to 1150 °C.
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Affiliation(s)
- Tiankai Yao
- Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, Troy, NY, 12180, USA
| | - Guoqing Xin
- Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, Troy, NY, 12180, USA
| | - Spencer Michael Scott
- Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, Troy, NY, 12180, USA
| | - Bowen Gong
- Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, Troy, NY, 12180, USA
| | - Jie Lian
- Department of Mechanical, Aerospace and Nuclear Engineering, Rensselaer Polytechnic Institute, Troy, NY, 12180, USA.
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Metzger KE, Knight TW, Roberts E, Huang X. Determination of mechanical behavior of U3Si2 nuclear fuel by microindentation method. Progress in Nuclear Energy 2017. [DOI: 10.1016/j.pnucene.2017.05.007] [Citation(s) in RCA: 20] [Impact Index Per Article: 2.9] [Reference Citation Analysis] [What about the content of this article? (0)] [Track Full Text] [Subscribe] [Scholar Register] [Indexed: 10/19/2022]
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