Browse > Article
http://dx.doi.org/10.1016/j.net.2021.09.034

Densification of matrix graphite for spherical fuel elements used in molten salt reactor via addition of green pitch coke  

He, Zhao (Key Laboratory of Carbon Materials, Institute of Coal Chemistry, Chinese Academy of Sciences)
Zhao, Hongchao (Key Laboratory of Carbon Materials, Institute of Coal Chemistry, Chinese Academy of Sciences)
Song, Jinliang (Shanghai Institute of Applied Physics, Chinese Academy of Sciences)
Guo, Xiaohui (Key Laboratory of Carbon Materials, Institute of Coal Chemistry, Chinese Academy of Sciences)
Liu, Zhanjun (Key Laboratory of Carbon Materials, Institute of Coal Chemistry, Chinese Academy of Sciences)
Zhong, Yajuan (Shanghai Institute of Applied Physics, Chinese Academy of Sciences)
Marrow, T. James (Department of Materials, University of Oxford)
Publication Information
Nuclear Engineering and Technology / v.54, no.4, 2022 , pp. 1161-1166 More about this Journal
Abstract
Green pitch coke with an average particle size of 2 mm was adopted as densifier and added to the raw materials of conventional A3-3 matrix graphite (MG) to prepare modified A3-3 matrix graphite (MMG) by the quasi-isostatic molding method. The structure, mechanical and thermal properties were assessed. Compared with MG, MMG had a more compact structure, and exhibited improved properties of higher mechanical strength, higher thermal conductivity and better molten salt barrier performance. Notably, under the same infiltration pressure of 5 atm, the fluoride salt occupation of MMG was only 0.26 wt%, whereas it was 15.82 wt% for MG. The densification effect of green pitch coke endowed MMG with improved properties for potential use in the spherical fuel elements of molten salt reactor.
Keywords
Green pitch coke; Matrix graphite; Molten salt reactor; Molten salt barrier property;
Citations & Related Records
Times Cited By KSCI : 3  (Citation Analysis)
연도 인용수 순위
1 M.W. Rosenthal, P.N. Haubenreich, R.B. Briggs, The Development Status of Molten-Salt Breeder Reactors, Oak Ridge National Laboratory, ORNL-4812, USA, 1972, pp. 273-290.
2 J.L. Song, Y.L. Zhao, X.J. He, et al., Preparation of pyrolytic carbon coating on graphite for inhibiting liquid fluoride salt and Xe135 penetration for molten salt breeder reactor, J. Nucl. Mater. 456 (2015) 33-40.   DOI
3 Z. He, P. Lian, Y. Song, et al., Protecting nuclear graphite from liquid fluoride salt and oxidation by SiC coating derived from polycarbosilane, J. Eur. Ceram. Soc. 38 (2) (2018) 453-462.   DOI
4 S. Chakraborty, D. Debnath, A.R. Mallick, et al., Microscopic, mechanical and thermal properties of spark plasma sintered ZrB2 based composite containing polycarbosilane derived SiC, Int. J. Refract. Metals Hard Mater. 52 (2015) 176-182.   DOI
5 Y. Zhong, J. Zhang, J. Lin, et al., Mesocarbon microbead based graphite for spherical fuel element to inhibit the infiltration of liquid fluoride salt in molten salt reactor, J. Nucl. Mater. 490 (2017) 34-40.   DOI
6 Z. He, J. Song, P. Lian, et al., Excluding molten fluoride salt from nuclear graphite by SiC/glassy carbon composite coating, Nucl. Eng. Techno. 51 (5) (2019) 1390-1397.   DOI
7 Z. He, L. Gao, W. Qi, et al., Molten FLiNaK salt infiltration into degassed nuclear graphite under inert gas pressure, Carbon 84 (2015) 511-518.   DOI
8 A. Awasthi, Y.J. Bhatt, S.P. Garg, Measurement of contact angle in systems involving liquid metals, Meas. Sci. Technol. 7 (5) (1996) 753-757.   DOI
9 P. Lian, J. Song, Z. Liu, et al., Preparation of ultrafine-grain graphite by liquid dispersion technique for inhibiting the liquid fluoride salt infiltration, Carbon 102 (2016) 208-215.   DOI
10 Z. He, P. Lian, Y. Song, et al., Improving molten fluoride salt and Xe135 barrier property of nuclear graphite by phenolic resin impregnation process, J. Nucl. Mater. 499 (2018) 79-87.   DOI
11 S. Feng, L. Xu, L. Li, et al., Sealing nuclear graphite with pyrolytic carbon, J. Nucl. Mater. 441 (1-3) (2013) 449-454.   DOI
12 X.J. He, J.L. Song, J. Tan, et al., SiC coating: an alternative for the protection of nuclear graphite from liquid fluoride salt, J. Nucl. Mater. 448 (1-3) (2014) 1-3.   DOI
13 Advances in High Temperature Gas Cooled Reactor Fuel Technology, Vienna. IAEA-TECDOC-CD-1674, Austria, 2012.
14 H. Mccoy, R. Beatty, W. Cook, et al., New developments in materials for molten salt reactors, Nucl. Appl. Technol. 8 (2) (1970) 156-169.   DOI
15 S. Xu, F. Kang, Carbon and Graphite Materials in Nuclear Engineering, vol. 204, Tsinghua University Press, Beijing (China), 2010, pp. 234-242.
16 H.-X. Xu, J. Lin, Y.-J. Zhong, et al., Characterization of molten 2LiF-BeF2 salt impregnated into graphite matrix of fuel elements for thorium molten salt reactor, Nucl. Sci. Tech. 30 (5) (2019).
17 J. Uhlir, Chemistry and technology of molten salt reactors - history and perspectives, J. Nucl. Mater. 360 (1) (2007) 6-11.   DOI
18 H. Wang, L. Xu, Y. Zhong, et al., Mesocarbon microbead densified matrix graphite A3-3 for fuel elements in molten salt reactors, Nucl. Eng. Techno. 53 (5) (2021) 1569-1579.   DOI
19 H. Zhao, Z. He, Z. Liu, et al., Self-sintered nanopore-isotropic graphite derived from green pitch coke for application in molten salt nuclear reactor, Ann. Nucl. Energy 131 (2019) 412-416.   DOI
20 B.A. Frandsen, S.D. Nickerson, A.D. Clark, et al., The structure of molten FLiNaK, J. Nucl. Mater. 537 (2020).
21 ASTM C695-91, Standard Test Method for Compressive Strength of Carbon and Graphite, 2010.
22 V. Bernardet, S. Gomes, S. Delpeux, et al., Protection of nuclear graphite toward fluoride molten salt by glassy carbon deposit, J. Nucl. Mater. 384 (3) (2009) 292-302.   DOI
23 Z. Dai, Thorium molten salt reactor nuclear energy system (TMSR), Molten Salt Reactors and Thorium Energy (2017) 531-540.
24 X. Zhou, Z. Lu, J. Zhang, et al., Study on the comprehensive properties and microstructures of A3-3 matrix graphite related to the high temperature purification treatment, Sci. Technol. Nuc. Ins. 2018 (2018) 1-10.
25 R.E. Schulze, H.A. Schulze, W. Rind, Graphitic Matrix Materials for Spherical HTR Fuel Elements, Jul-Spez-167, 1982, pp. 3-5.
26 R.B. Briggs, P.R. Kasten, Molten-salt Reactor Program Semiannual Progress Report, Oak Ridge National Laboratory, ORNL-3419, USA, 1963, pp. 71-73.
27 P.R. Kasten, E.S. Bettis, W.H. Cook, et al., Graphite behavior and its effects on MSBR performance, Nucl. Eng. Des. 9 (2) (1969) 157-195.   DOI
28 X.J. He, J.L. Song, L. Xu, et al., Protection of nuclear graphite toward liquid fluoride salt by isotropic pyrolytic carbon coating, J. Nucl. Mater. 442 (1-3) (2013) 306-308.   DOI
29 Rene Blachard, Saint-Martin-le-Vinoux, Louis Bochirol, et al., Process for the Densification of Carbonaceous Bodies, U.S. Patent 3321327, 1967.
30 H.-c. Zhao, Z. He, X.-h. Guo, et al., Effect of the average grain size of green pitch coke on the microstructure and properties of self-sintered graphite blocks, N. Carbon Mater. 35 (2) (2020) 184-192.   DOI
31 Pore Size Distribution and Porosity of Solid Materials by Mercury Porosimetry and Gas Adsorption-Part 1: Mercury Porosimetry, 2005, p. 1. GB/T 21650.1-2008/ISO15901-1.
32 E.W. Washburn, The dynamics of capillary flow, Phys. Rev. 17 (3) (1921) 273-283.   DOI