| Review |
|
|
|
|
|
Development strategy of key materials technology for the high temperature gas-cooled reactor |
| SHI Li, ZHAO Jiaqing, LIU Bing, LI Xiaowei, LUO Xiaowei, ZHANG Zhengming, ZHANG Ping, SUN Libin, WU Xinxin |
| Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084, China |
|
|
|
|
Abstract The development of nuclear power systems in China requires that key materials for the reactor core and the main components be produced domestically. The key materials investigated in this study included the nuclear fuel, high temperature metals, nuclear graphite, pressure vessel materials, and high temperature gas-cooled reactor (HTGR) hydrogen production related materials. However, current domestic research and development and manufacturing capabilities require that some key HTGR materials still be purchased from abroad. Thus, China has sought to develop domestically produced key HTGR materials and technologies. This study analyzed the fundamental roles of various aspects of some key materials on the development of HTGR systems, including the content and scope, industry manufacturing chain, and characterization and application of key materials. The results indicate that research systems should be developed to support engineering development and industrial manufacturing of these key materials to support HTGR development. Future development schedules and suggestions are provided.
|
| Keywords
high temperature gas-cooled reactor (HTGR)
key materials
nuclear fuel
high temperature metal
nuclear graphite
materials related to hydrogen production
|
|
Issue Date: 16 April 2021
|
|
|
[1] ZHANG Z Y, YU S Y. Future HTGR developments in China after the criticality of the HTR-10[J]. Nuclear Engineering and Design, 2002, 218(1-3): 249-257.
[2] LIU R Z, LIU M L, CHANG J X, et al. An improved design of TRISO particle with porous SiC inner layer by fluidized bed-chemical vapor deposition [J]. Journal of Nuclear Materials, 2015, 467: 917-926.
[3] PETTI D A, BUONGIORNO J, MAKI J T, et al. Key differences in the fabrication, irradiation and high temperature accident testing of US and German TRISO-coated particle fuel, and their implications on fuel performance [J]. Nuclear Engineering and Design, 2003, 222(2-3): 281-297.
[4] KNOL S, DE GROOT S, SALAM R V, et al. HTR-PM fuel pebble irradiation qualification in the high flux reactor in Petten [J]. Nuclear Engineering and Design, 2018, 329: 82-88.
[5] FREIS D, EL ABJANI A, CORIC D, et al. Burn-up determination and accident testing of HTR-PM fuel elements irradiated in the HFR Petten [J]. Nuclear Engineering and Design, 2020, 357: 110414.
[6] MINATO K, OGAWA T, FUKUDA K, et al. Fission product release from ZrC-coated fuel particles during post-irradiation heating at 1800 and 2000℃ [J]. Journal of Nuclear Materials, 1997, 249(2-3): 142-149.
[7] 刘超, 刘兵, 邵友林, 等. 高温气冷堆包覆燃料颗粒ZrC涂层的研究进展[J]. 稀有金属材料与工程, 2007, 36(S3): 674-678. LIU C, LIU B, SHAO Y L, et al. The research progress of ZrC coating layer of fuel particles for HTGR at home and abroad [J]. Rare Metal Materials and Engineering, 2007, 36(S3): 674-678. (in Chinese)
[8] FORSBERG C W, TERRANI K A, SNEAD L L, et al. Fluoride-salt-cooled high-temperature reactor (FHR) with silicon-carbide-matrix coated-particle fuel [J]. Transactions of the American Nuclear Society, 2012, 107: 907-910.
[9] FIELDING R, MEYER M, JUE J F, et al. Gas-cooled fast reactor fuel fabrication [J]. Journal of Nuclear Materials, 2007, 371(1-3): 243-249.
[10] ŞAHIN S, SEFIDVASH F. The fixed bed nuclear reactor concept [J]. Energy Conversion and Management, 2008, 49(7): 1902-1909.
[11] YUEH K, CARPENTER D, FEINROTH H. Clad in clay [J]. Nuclear Engineering International, 2010, 55(666): 14-16.(查阅所有网上资料,期不确定是否正确,请联系作者确认)
[12] 赵熹. 超高温气冷堆结构材料力学性能及寿命预测研究[D]. 北京: 清华大学, 2017.(查阅所有网上资料,未找到本条文献信息,请联系作者确认) ZHAO X. Research on mechanical properties and life prediction of structural materials for very high temperature reactor [D]. Beijing: Tsinghua University, 2017. (in Chinese)
[13] REN W J, SWINDEMAN R. A review paper on aging effects in alloy 617 for Gen IV nuclear reactor applications [J]. Journal of Pressure Vessel Technology, 2009, 131(2): 024002-15.
[14] The American Society of Mechanical Engineers. ASME boiler & pressure vessel code, section Ⅲ-Rules for construction of nuclear facility components, subsection NH [M]. ASME Press, 2019.(查阅所有网上资料,未找到本条文献信息,请联系作者确认)
[15] NATESAN K, PUROHIT A, TAM S W. Materials behavior in HTGR environments [R]. Argonne National Laboratory Report, ANL-02/37, 2003.(查阅所有网上资料,文献类型不确定是否正确,请联系作者确认)
[16] 徐世江. 核工程中的石墨和炭素材料(第三讲) [J]. 炭素技术, 2000(3): 44-48. XU S J. Graphite and carbonaceous material in nuclear engineering [J]. Carbon Techniques, 2000(3): 44-48. (in Chinese)
[17] 徐世江. 核工程中的石墨和炭素材料(第五讲) [J]. 炭素技术, 2000(5): 42-46. XU S J. Graphite and carbonaceous material in nuclear engineering [J]. Carbon Techniques, 2000(5): 42-46. (in Chinese)
[18] NEMETH N N, BRATTON R L. Overview of statistical models of fracture for nonirradiated nuclear-graphite components [J]. Nuclear Engineering and Design, 2010, 240(1): 1-29.
[19] HAAG G. Properties of ATR-2E graphite and property changes due to fast neutron irradiation. Juel-4183[R]. Forschungszentrum Jülich GmbH Zentralbibliothek, 2005.(查阅所有网上资料,未找到出版地信息,请联系作者确认)
[20] BURCHELL T, BRATTON R, WINDES W. NGNP graphite selection and acquisition strategy: ORNL/TM-2007/153[R]. Oak Ridge: Oak Ridge National Laboratory, 2007.
[21] 胡玉琴, 孙立斌, 王洪涛, 等. 高温气冷堆工程验证用国产石墨的强度实验研究[J]. 科技导报, 2012, 30(21): 45-50. HU Y Q, SUN L B, WANG H T, et al. Domestic graphite strength test for HTR engineering verification [J]. Science & Technology Review, 2012, 30(21): 45-50. (in Chinese)
[22] 何西扣, 刘正东, 赵德利, 等. 中国核压力容器用钢及其制造技术进展[J]. 中国材料进展, 2020, 39(7-8): 509-518, 557. HE X K, LIU Z D, ZHAO D L, et al. Progress on nuclear pressure vessel steels and its manufacturing technology in China [J]. Materials China, 2020, 39(7-8): 509-518, 557. (in Chinese)
[23] PINSKY R, SABHARWALL P, HARTVIGSEN J, et al. Comparative review of hydrogen production technologies for nuclear hybrid energy systems [J]. Progress in Nuclear Energy, 2020, 123: 103317.
[24] ZHANG P, WANG L J, CHEN S Z, et al. Progress of nuclear hydrogen production through the iodine-sulfur process in China [J]. Renewable and Sustainable Energy Reviews, 2018, 81: 1802-1812. |
|
Viewed |
|
|
|
Full text
|
|
|
|
|
Abstract
|
|
|
|
|
Cited |
|
|
|
|
| |
Shared |
|
|
|
|
| |
Discussed |
|
|
|
|
2021,
Vol. 61
