Abstract
The modular high-temperature gas-cooled reactor (MHTGR) is known to be inherently safe because the fuel cannot fail even if all engineered safety systems, including decay heat removal and reactivity control, fail. In the event of a loss of cooling accident, even with failure of passive convective systems, core afterheat will safely be conducted to ground. However, it is also known that the MHTGR core is physically big when compared with current commercial reactors, the tri-isotropic (TRISO)-coated particle fuel is expensive and limits the uranium loading, and the reactor vessel is very large despite modest power output. Earlier economic evaluations indicate the truly inherently safe MHTGR has a weakness as a commercial product due to high cost. This study investigates the neutronic feasibility of a fuel alternative similar to the conventional pellet-type fuel, which can simplify the fuel fabrication process and reduce the enrichment of the fuel, while maintaining the accident-tolerant fuel (ATF) features. The neutronics performance of such a fuel concept has been evaluated for the amount of fuel loading, excess reactivity, and fuel cycle length, followed by thermal–mechanical performance and inherent safety analyses of such a design. The preliminary investigations have shown that it is feasible to construct a fuel cycle using uranium carbide (UC) fuel of ~5 wt% enrichment.
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References
M. LaBar and A. Shenoy, W. Simon, E. Campbell, “The Gas Turbine-Modular Helium Reactor,” Nuclear News, Vol. 46, pp. 2, 2003
F. Goldner, “Development Strategy for Advanced LWR Fuels with Enhanced Accident Tolerance,” Nuclear Energy Advisory Committee Meeting, Washington D.C., U.S.A., June 12, 2012
D.M. Carpenter, “Assessment of Innovative Fuel Designs for High Performance Light Water Reactors,” M.S. Thesis, Massachusetts Institute of Technology, 2006
K. Barrett, S. Bragg-Sitton, and D. Galicki, “Advanced LWR Nuclear Fuel Cladding System Development Trade-off Study,” INL/EXT-12-27090, Idaho National Laboratory, September 2012
S. Ray, E. Lahoda, and F. Franceschini, “Assessment of Different Materials for Meeting the Requirement of Future Fuel Designs,” 2012 Reactor Fuel Performance Meeting, Manchester, UK, September 2–6, 2012
R. A. Rucker, “450 MW(t) MHTGR Core Nuclear Design,” DOE-HTGR-90237, General Atomics, September 1993
F.A. Silady, W.A. Simon, “Innovative Safety Features of the Modular HTGR,” GA-A20789, General Atomics, April 1992
D. Chung, “Materials for Thermal Conduction,” Applied Thermal Engineering, 21, pp. 1593–1605, 2001
S. Nishigaki, “Utilization of BeO in High Temperature Gas Cooled Reactor,” J. of the Atomic Energy Society of Japan, Vol. 6, pp. 583–593, 1964
S.K. Kim et al., “Economic Viability of BeO-UO2 Fuel Burnup Extension,” Nuclear Engineering and Technology, Vol. 43, pp. 141–148, 2011
“MCNP – A General Monte Carlo N-Particle Transport Core, Version 5,” LA-UR-03-1987, Los Alamos National Laboratory, April 2003
M. Suzuki and H. Saitou, “Light Water Fuel Analysis Code FEMAXI-6 (Ver. 1),” JAEA-Data/Code 2005-003, Japan Atomic Energy Agency, 2005
H. Choi, “Evaluation of Carbide Fuel Property Data and Models,” Reactor Fuel Performance Meeting, TopFuel 2015, Zurich, Switzerland, Sept. 13–17, 2015
H. Choi, “LWR Fuel Behavior with SiC Cladding,” 2013 ANS Annual Meeting, Atlanta, USA, June 16–20, 2013
D.L. Hagman and G.A. Reyman, “MATPRO-Version 11, A Handbook of Materials Properties for Use in the Analysis of Light Water Reactor Fuel Rod Behavior,” NUREG/CR-0497, TREE-1280 Rev.3, 1979
T. Preusser, “Modeling of Carbide Fuel Rods,” Nuclear Technology, 57, pp. 343–371, 1982
T. Nakajima et al., “FEMAXI-III: A Computer Code for the Analysis of Thermal and Mechanical Behavior of Fuel Rods,” JAERI-1298, Japan Atomic Energy Research Institute, 1985
W. Dienst, “Swelling, Densification and Creep of (U,Pu)C Fuel Under Irradiation,” J. Nuclear Materials, 124, pp. 153–158, 1984
L.L. Snead et al., “Handbook of SiC Properties for Fuel Performance Modeling,” J. Nuclear Materials, 371, pp. 329–377, 2007
R. Boonstra, “TAC2D – A General Purpose Two-Dimensional Heat Transfer Computer Code,” GA-A14032, General Atomics, July 1976
H. Sato et al., “Thermal analysis of heated cylinder simulating nuclear reactor during loss of coolant accident,” J. Nuclear Science and Technology, 51, pp. 1317–1323, 2014
K. Kunitomi et al., “Reactor core design of Gas Turbine High Temperature reactor 300,” Nuclear Engineering and Design, 230, pp. 349–366, 2004
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This work was supported by General Atomics internal funding.
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Choi, H., Schleicher, R., Choi, M. (2017). Physics Analysis of Alternative Fuel Options for HTGR. In: Jiang, H. (eds) Proceedings of The 20th Pacific Basin Nuclear Conference. PBNC 2016. Springer, Singapore. https://doi.org/10.1007/978-981-10-2317-0_76
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DOI: https://doi.org/10.1007/978-981-10-2317-0_76
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