Acknowledgement
This project was supported by the National Natural Science Foundation of China, No. 11975219, and the Scientific Research Program for Young Talent Elite Project of the China National Nuclear Corporation (CNNC2019YTEP-NPIC01). The authors are also thankful for the support of the Science and Technology on Reactor System Design Technology Laboratory, Nuclear Power Institute of China.
References
- H. Yu, Y. Ma, Z. Zhang, X. Chai, Initiation and development of heat pipe cooled reactor, Nucl. Power Eng. 40 (2019) 1-8.
- B.H. Yan, C. Wang, L.G. Li, The technology of micro heat pipe cooled reactor: a review, Ann. Nucl. Energy 135 (2020) 106948. https://doi.org/10.1016/j.anucene.2019.106948
- J.W. Sterbentz, J.E. Werner, M.G. McKellar, A.J. Hummel, J.C. Kennedy, R.N. Wright, J.M. Biersdorf, Special purpose nuclear reactor (5 MW) for reliable power at remote sites assessment report, in: Idaho National Lab (INL), 2017. Idaho Falls, ID (United States).
- Y. Ma, M. Liu, B. Xie, W. Han, H. Yu, S. Huang, X. Chai, Y. Liu, Z. Zhang, Neutronic and thermal-mechanical coupling analyses in a solid-state reactor using Monte Carlo and finite element methods, Ann. Nucl. Energy 151 (2021) 107923. https://doi.org/10.1016/j.anucene.2020.107923
- H. Cao, G. Wang, The research on the heat transfer of a solid-core nuclear reactor cooled by heat pipe through a numerical simulation, considering the assembly gaps, Ann. Nucl. Energy 130 (2019) 431-439. https://doi.org/10.1016/j.anucene.2019.03.013
- Y. Ma, E. Chen, H. Yu, R. Zhong, J. Deng, X. Chai, S. Huang, S. Ding, Z. Zhang, Heat pipe failure accident analysis in megawatt heat pipe cooled reactor, Ann. Nucl. Energy 149 (2020) 107755. https://doi.org/10.1016/j.anucene.2020.107755
- P. Hu, L. Chen, L. Wang, L. Zhang, Steady-state thermal analysis of a fuel module in the heat pipe-cooled reactor (in Chinese), Modern Appl. Phys. 4 (2013) 375-378.
- Y.S. Kim, G. Hofman, Fission product induced swelling of U-Mo alloy fuel, J. Nucl. Mater. 419 (2011) 291-301. https://doi.org/10.1016/j.jnucmat.2011.08.018
- M. Meyer, J. Gan, J. Jue, D. Keiser, E. Perez, A. Robinson, D. Wachs, N. Woolstenhulme, G. Hofman, Y. Kim, Irradiation performance of U-Mo monolithic fuel, Nucl. Eng. Technol. 46 (2014) 169-182. https://doi.org/10.5516/NET.07.2014.706
- K. Une, M. Imamura, M. Amaya, Y. Korei, Fuel oxidation and irradiation behaviors of defective BWR fuel rods, J. Nucl. Mater. 223 (1995) 40-50. https://doi.org/10.1016/0022-3115(94)00693-8
- C. Tang, Y. Jiao, Y. Li, Y. Zhou, H. Pang, Preliminary research on the irradiation-thermal-mechanical coupling behavior simulation method of FCM fuel, Int. J. Adv. Nuclear Reactor Design Technol. 1 (2019) 51-56. https://doi.org/10.1016/j.jandt.2019.10.002
- Y. Zhao, X. Gong, S. Ding, Simulation of the irradiation-induced thermo-mechanical behaviors evolution in monolithic U-Mo/Zr fuel plates under a heterogeneous irradiation condition, Nucl. Eng. Des. 285 (2015) 84-97. https://doi.org/10.1016/j.nucengdes.2014.12.030
- P.R. Mcclure, D.I. Poston, V.R. Dasari, R.S. Reid, Design of megawatt power level heat pipe reactors, in: Los Alamos National Lab (LANL), 2015. Los Alamos, NM (United States).
- D.I. Poston, The Heatpipe-Operated Mars Exploration Reactor (HOMER), AIP, American Institute of Physics, United States, 2001.
- Y. Ma, C. Tian, H. Yu, R. Zhong, Z. Zhang, S. Huang, J. Deng, X. Chai, Y. Yang, Transient heat pipe failure accident analysis of a megawatt heat pipe cooled reactor, Prog. Nucl. Energy 140 (2021) 103904. https://doi.org/10.1016/j.pnucene.2021.103904
- Z.J. Zuo, A. Faghri, A network thermodynamic analysis of the heat pipe, Int. J. Heat Mass Tran. 41 (1998) 1473-1484. https://doi.org/10.1016/S0017-9310(97)00220-2
- Y. Ma, R. Zhong, Z. Zhang, X. Chai, S. Huang, H. Yu, A transient and steady-state network model forannular-wick heat pipes in continuum flow pattern, in: 27th International Conference on Nuclear Engineering, ICONE 2019, May 19-24, 2019, Tsukuba, Ibaraki, Japan, 2019.
- K. Wang, Z. Li, D. She, J.G. Liang, Q. Xu, Y. Qiu, J. Yu, J. Sun, X. Fan, G. Yu, RMC - a Monte Carlo code for reactor core analysis, Ann. Nucl. Energy 82 (2015) 121-129. https://doi.org/10.1016/j.anucene.2014.08.048
- A. Ross, R. Stoute, Heat Transfer Coefficient between UO 2 and Zircaloy-2, Atomic Energy of Canada Limited, Chalk River, Ontario, 1962.
- K. Geelhood, W. Luscher, P. Raynaud, I. Porter, A computer code for the calculation of steady-state, thermal-mechanical behavior of oxide fuel rods for high burnup, in: Pacific Northwest, National Laboratory, Richland, Washington, 2015.
- H. Yu, W. Tian, Z. Yang, G.H. Su, S. Qiu, Development of fuel rod behavior analysis code (FROBA) and its application to AP1000, Ann. Nucl. Energy 50 (2012) 8-17. https://doi.org/10.1016/j.anucene.2012.06.010
- Y. Rashid, R. Dunham, R. Montgomery, Fuel Analysis and Licensing Code: FALCON MOD01, EPRI Report, 2004.
- D.L. Hagrman, G.A. Reymann, MATPRO-Version 11: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior, in: Idaho National Engineering Lab, 1979. Idaho Falls (USA).
- B. Mihaila, M. Stan, J. Ramirez, A. Zubelewicz, P. Cristea, Simulations of coupled heat transport, oxygen diffusion, and thermal expansion in UO2 nuclear fuel elements, J. Nucl. Mater. 394 (2009) 182-189. https://doi.org/10.1016/j.jnucmat.2009.09.007
- P. Lucuta, I. Hastings, A pragmatic approach to modelling thermal conductivity of irradiated UO2 fuel: review and recommendations, J. Nucl. Mater. 232 (1996) 166-180. https://doi.org/10.1016/S0022-3115(96)00404-7
- K.C. Mills, Y. Su, Z. Li, R.F. Brooks, Equations for the calculation of the thermophysical properties of stainless steel, ISIJ Int. 44 (2004) 1661-1668. https://doi.org/10.2355/isijinternational.44.1661
- W.G. Luscher, K.J. Geelhood, Material property correlations: comparisons between FRAPCON-3.4, FRAPTRAN 1.4, and MATPRO, in: Pacific Northwest National Lab.(PNNL), 2010. Richland, WA (United States).
- P. McClure, D. Poston, D. Rao, R. Reid, Design of megawatt power level heat pipe reactors, in: Los Alamos National Lab.(LANL), 2015. Los Alamos, NM (United States).
- M.R. Eslami, R.B. Hetnarski, J. Ignaczak, N. Noda, N. Sumi, Y. Tanigawa, Theory of Elasticity and Thermal Stresses, Springer, 2013.
- A. Booth, A method of calculating fission gas diffusion from UO 2 fuel and its application to the X-2-f loop test, in: Atomic Energy of Canada Limited, 1957. Chalk River, Ontario.
- L. Bernard, J. Jacoud, P. Vesco, An efficient model for the analysis of fission gas release, J. Nucl. Mater. 302 (2002) 125-134. https://doi.org/10.1016/S0022-3115(02)00793-6
- P. Van Uffelen, M. Suzuki, Oxide fuel performance modeling and simulations, in: R.J.M. Konings (Ed.), Comprehensive Nuclear Materials, Elsevier, Oxford, 2012, pp. 535-577.
- J. Rest, M. Cooper, J. Spino, J. Turnbull, P. Van Uffelen, C. Walker, Fission gas release from UO2 nuclear fuel: a review, J. Nucl. Mater. 513 (2019) 310-345. https://doi.org/10.1016/j.jnucmat.2018.08.019
- P. Li, H. Dong, Y. Xia, X. Hao, S. Wang, L. Pan, J. Zhou, Inhomogeneous interface structure and mechanical properties of rotary friction welded TC4 titanium alloy/316L stainless steel joints, J. Manuf. Process. 33 (2018) 54-63. https://doi.org/10.1016/j.jmapro.2018.05.001