DOI QR코드

DOI QR Code

A review on thermohydraulic and mechanical-physical properties of SiC, FeCrAl and Ti3SiC2 for ATF cladding

  • Qiu, Bowen (Shaanxi Key Laboratory of Advanced Nuclear Energy and Technology, Shaanxi Engineering Research Center of Advanced Nuclear Energy, State Key Laboratory of Multiphase Flow in Power Engineering, Xi'an Jiaotong University) ;
  • Wang, Jun (University of Wisconsin-Madison) ;
  • Deng, Yangbin (College of Physics and Optoelectronic Engineering, Shenzhen University) ;
  • Wang, Mingjun (Shaanxi Key Laboratory of Advanced Nuclear Energy and Technology, Shaanxi Engineering Research Center of Advanced Nuclear Energy, State Key Laboratory of Multiphase Flow in Power Engineering, Xi'an Jiaotong University) ;
  • Wu, Yingwei (Shaanxi Key Laboratory of Advanced Nuclear Energy and Technology, Shaanxi Engineering Research Center of Advanced Nuclear Energy, State Key Laboratory of Multiphase Flow in Power Engineering, Xi'an Jiaotong University) ;
  • Qiu, S.Z. (Shaanxi Key Laboratory of Advanced Nuclear Energy and Technology, Shaanxi Engineering Research Center of Advanced Nuclear Energy, State Key Laboratory of Multiphase Flow in Power Engineering, Xi'an Jiaotong University)
  • Received : 2019.04.19
  • Accepted : 2019.07.29
  • Published : 2020.01.25

Abstract

At present, the Department of Energy (DOE) in Unite State are directing the efforts of developing accident tolerant fuel (ATF) technology. As the first barrier of nuclear fuel system, the material selection of fuel rod cladding for ATFs is a basic but very significant issue for the development of this concept. The advanced cladding is attractive for providing much stronger oxidation resistance and better in-pile behavior under sever accident conditions (such as SBO, LOCA) for giving more coping time and, of course, at least an equivalent performance under normal condition. In recent years, many researches on in-plie or out-pile physical properties of some suggested cladding materials have been conducted to solve this material selection problem. Base on published literatures, this paper introduced relevant research backgrounds, objectives, research institutions and their progresses on several main potential claddings include triplex SiC, FeCrAl and MAX phase material Ti3SiC2. The physical properties of these claddings for their application in ATF area are also reviewed in thermohydraulic and mechanical view for better understanding and simulating the behaviors of these new claddings. While most of important data are available from publications, there are still many relevant properties are lacking for the evaluations.

Keywords

References

  1. F. Kamil, Fukushima nuclear accident, Decouverte (Paris) 1 (373) (2011) 28-31.
  2. J. Carmack, F. Goldner, S.M. Bragg-Sitton, L.L. Snead, Overview of the US DOE accident tolerant fuel development program, in: Proc. 2013 LWR Fuel Performance Meeting/TopFuel, 2013, pp. 15-19.
  3. L. Soffer, S. Burson, C. Ferrell, R. Lee, J. Ridgely, Accident Source Terms for Light-Water Nuclear Power Plants. NUREG-1465 6, U.S.NRC, Washington, DC, United States, 1995.
  4. F. Goldner, Development Strategy for Advanced LWR Fuels with Enhanced Accident Tolerant, Enhanced Accident Tolerant LWR Fuels National Metrics Workshop 2012, 2012 (Germantown, MD, United States).
  5. S. Bragg-Sitton, Development of advanced accident-tolerant fuels for commercial LWRs, Nucl. News 57 (2014) 83.
  6. S. Chu, A. Majumdar, Opportunities and challenges for a sustainable energy future, Nature 488 (2012) 294-303. https://doi.org/10.1038/nature11475
  7. K.R. Gurney, D.L. Mendoza, Y. Zhou, M.L. Fischer, C.C. Miller, S. Geethakumar, S. dela Rue du Can, High resolution fossil fuel combustion CO2 emission fluxes for the United States, Environ. Sci. Technol. 43 (2009) 5535-5541. https://doi.org/10.1021/es900806c
  8. Y.-H. Koo, J.-H. Yang, J.-Y. Park, K.-S. Kim, H.-G. Kim, D.-J. Kim, Y.-I. Jung, K.-W. Song, KAERI's development of LWR accident-tolerant fuel, Nucl. Technol. 186 (2014) 295-304. https://doi.org/10.13182/NT13-89
  9. B.A. Pint, K.A. Terrani, M.P. Brady, T. Cheng, J.R. Keiser, High temperature oxidation of fuel cladding candidate materials in steamehydrogen environments, J. Nucl. Mater. 440 (2013) 420-427. https://doi.org/10.1016/j.jnucmat.2013.05.047
  10. K. Yueh, K.A. Terrani, Silicon carbide composite for light water reactor fuel assembly applications, J. Nucl. Mater. 448 (2014) 380-388. https://doi.org/10.1016/j.jnucmat.2013.12.004
  11. L.J. Ott, K.R. Robb, D. Wang, Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions, J. Nucl. Mater. 448 (2014) 520-533. https://doi.org/10.1016/j.jnucmat.2013.09.052
  12. N.M. George, K. Terrani, J. Powers, A. Worrall, I. Maldonado, Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors, Ann. Nucl. Energy 75 (2015b) 703-712. https://doi.org/10.1016/j.anucene.2014.09.005
  13. F. Goldner, Development Strategy for Advanced LWR Fuels with Enhanced Accident Tolerant, Enhanced Accident Tolerant LWR Fuels National Metrics Workshop 2012,Germantown, MD, United States, 2012.
  14. J. Wang, M. Mccabe, L. Wu, et al., Accident tolerant clad material modeling by MELCOR: benchmark for SURRY short term station black out, Nucl. Eng. Des. 313 (2017a) 458-469. https://doi.org/10.1016/j.nucengdes.2017.01.002
  15. J. Wang, H.J. Jo, M.L. Corradini, et al., "Potential recovery actions from a severe accident in a PWR: MELCOR analysis of a station blackout scenario", Nucl. Technol. 204 (1) (2018) 1-14. https://doi.org/10.1080/00295450.2018.1464838
  16. J. Wang, H.J. Jo, M.L. Corradini, Accident Tolerant Fuel (ATF) Coating and Cladding Thermal Hydraulic Properties Evaluation by MELCOR YU 1.8.6: Benchmark for SURRY Short Term Station Black Out. 2018 ANS Annual Meeting, 2018. June 17-21, Philadelphia, PA, United States.
  17. J. Wang, A. Gurgen, M.L. Corradini, et al., Accident tolerant fuel benchmark calculation by MELCOR and TRACE for a simplified generic pressure water reactor. 2018, in: International Congress on Advances in Nuclear Power Plants, 2018. April.8-11, Charlotte, NC, United States.
  18. M.N. Chong, S. Lei, B. Jin, C. Saint, C.W. Chow, Optimisation of an annular photoreactor process for degradation of Congo Red using a newly synthesized titania impregnated kaolinite nano-photocatalyst, Separ. Purif. Technol. 67 (2009) 355-363. https://doi.org/10.1016/j.seppur.2009.04.001
  19. P. Hofmann, S.J. Hagen, V. Noack, G. Schanz, L.K. Sepold, Chemical-physical behavior of light water reactor core components tested under severe reactor accident conditions in the CORA facility, Nucl. Technol. 118 (1997) 200-224. https://doi.org/10.13182/NT118-200
  20. D. Squarer, A. Pieczynski, L. Hochreiter, Effect of debris bed pressure, particle size, and distribution on degraded nuclear reactor core coolability, Nucl. Sci. Eng. 80 (1982) 2-13. https://doi.org/10.13182/NSE82-A21399
  21. S.J. Zinkle, K.A. Terrani, J.C. Gehin, L.J. Ott, L.L. Snead, Accident tolerant fuels for LWRs: a perspective, J. Nucl. Mater. 448 (2014) 374-379. https://doi.org/10.1016/j.jnucmat.2013.12.005
  22. K.A. Terrani, S.J. Zinkle, L.L. Snead, Advanced oxidation-resistant iron-based alloys for LWR fuel cladding, J. Nucl. Mater. 448 (2014b) 420-435. https://doi.org/10.1016/j.jnucmat.2013.06.041
  23. S.J. Zinkle, G. Was, Materials challenges in nuclear energy, Acta Mater. 61 (2013) 735-758. https://doi.org/10.1016/j.actamat.2012.11.004
  24. G. Gonzalez-Doncel, O. Sherby, High temperature creep behavior of metal matri Aluminum SiC composites, Acta Metall. Mater. 41 (1993) 2797-2805. https://doi.org/10.1016/0956-7151(93)90094-9
  25. A. Evans, C. Padgett, R. Davidge, Strength of pyrolytic SiC coatings of fuel particles for high- temperature gas- cooled reactors, J. Am. Ceram. Soc. 56 (1973) 36-41. https://doi.org/10.1111/j.1151-2916.1973.tb12347.x
  26. M. Ben-Belgacem, V. Richet, K.A. Terrani, Y. Katoh, L.L. Snead, Thermomechanical analysis of LWR SiC/SiC composite cladding, J. Nucl. Mater. 447 (2014) 125-142. https://doi.org/10.1016/j.jnucmat.2014.01.006
  27. M. Baba, Fukushima accident: what happened? Radiat. Meas. 55 (2013) 17-21. https://doi.org/10.1016/j.radmeas.2013.01.013
  28. P.C. Burns, R.C. Ewing, A. Navrotsky, Nuclear fuel in a reactor accident, Science (Washington, D.C.) 335 (2012) 1184-1188. https://doi.org/10.1126/science.1211285
  29. N. Kinoshita, K. Sueki, K. Sasa, J.-i. Kitagawa, S. Ikarashi, T. Nishimura, Y.-S. Wong, Y. Satou, K. Handa, T. Takahashi, Assessment of individual radionuclide distributions from the Fukushima nuclear accident covering centraleast Japan, Proc. Natl. Acad. Sci. 108 (2011) 19526-19529. https://doi.org/10.1073/pnas.1111724108
  30. J.M. Schwantes, C.R. Orton, R.A. Clark, Analysis of a nuclear accident: fission and activation product releases from the Fukushima Daiichi nuclear facility as remote indicators of source identification, extent of release, and state of damaged spent nuclear fuel, Environ. Sci. Technol. 46 (2012) 8621-8627. https://doi.org/10.1021/es300556m
  31. B.B. Wittneben, The impact of the Fukushima nuclear accident on European energy policy, Environ. Sci. Policy 15 (2012) 1-3. https://doi.org/10.1016/j.envsci.2011.09.002
  32. K. Barrett, S. Bragg-Sitton, D. Galicki, Advanced LWR Nuclear Fuel Cladding System Development Trade-Off Study. INL/EXT-12-27090, INL External Report, 2012 (Idaho Falls, ID, United States).
  33. B.A. Pint, K.A. Terrani, Y. Yamamoto, L.L. Snead, Material selection for accident tolerant fuel cladding, Metall. Mater. Trans. A 2 (2015) 190-196.
  34. J.-Y. Park, I.-H. Kim, Y.-I. Jung, H.-G. Kim, D.-J. Park, B.-K. Choi, High temperature steam oxidation of Al 3 Ti-based alloys for the oxidation-resistant surface layer on Zr fuel claddings, J. Nucl. Mater. 437 (2013b) 75-80. https://doi.org/10.1016/j.jnucmat.2013.01.338
  35. S. Ray, E. Lahoda, F. Franceschini, Assessment of Different Materials for Meeting the Requirement of Future Fuel Designs, Reactor Fuel Performance Meeting, 2012, pp. 2-6. Paper A.
  36. M.C. Teague, B.S. Fromm, M.R. Tonks, D.P. Field, Using coupled mesoscale experiments and simulations to investigate high burn-up oxide fuel thermal conductivity, JOM (J. Occup. Med.) 66 (2014) 2569-2577.
  37. M. Tonks, D. Schwen, Y. Zhang, P. Chakraborty, X. Bai, B. Fromm, J. Yu, M. Teague, D. Andersson, Assessment of MARMOT: A Mesoscale Fuel Performance Code, Idaho National Laboratory (INL), Idaho Falls, ID, United States, 2015.
  38. K. Metzger, T. Knight, R. Williamson, Model of $U_3Si_2$ fuel system using BISON fuel code. Proceedings of the International Congress on Advances in Nuclear Power PlantseICAPP 2014, 2014 (Charlotte, NC, United States).
  39. J. Carmack, L. Braase, C. Papesch, D. Hurley, M. Tonks, Y. Zhang, K. Gofryk, J. Harp, R. Fielding, C. Knight, Thermal Properties Measurement Report, Idaho National Laboratory (INL), Idaho Falls, United States, 2015.
  40. K. Barrett, K. Ellis, C. Glass, G. Roth, M. Teague, J. Johns, Critical processes and parameters in the development of Accident Tolerant Fuel drop-in capsule irradiation tests, Nucl. Eng. Des. 294 (2015) 38-51. https://doi.org/10.1016/j.nucengdes.2015.07.074
  41. B. Qiu, et al., A comparative study on preliminary performance evaluation of ATFs under normal and accident conditions with FRAP-ATF code, Prog. Nucl. Energy 105 (2018) 51-60. https://doi.org/10.1016/j.pnucene.2017.12.010
  42. Y.B. Deng, Y.W. Wu, B.W. Qiu, et al., Development of a new pellet-clad mechanical interaction (PCMI) model and its application in ATFs, Ann. Nucl. Energy 104 (2017) 146-156. https://doi.org/10.1016/j.anucene.2017.02.022
  43. N.R. Brown, A. Aronson, M. Todosow, R. Brito, K.J. McClellan, Neutronic performance of uranium nitride composite fuels in a PWR, Nucl. Eng. Des. 275 (2014) 393-407. https://doi.org/10.1016/j.nucengdes.2014.04.040
  44. V. Mehta, J.S. Cooper, Review and analysis of PEM fuel cell design and manufacturing, J. Power Sources 114 (2003) 32-53. https://doi.org/10.1016/S0378-7753(02)00542-6
  45. M. Farmer, L. Leibowitz, K.A. Terrani, K.R. Robb, Scoping assessments of ATF impact on late-stage accident progression including molten coreeconcrete interaction, J. Nucl. Mater. 448 (2014) 534-540. https://doi.org/10.1016/j.jnucmat.2013.12.022
  46. B.A. Pint, S. Dryepondt, K.A. Unocic, D.T. Hoelzer, Development of ODS FeCrAl for compatibility in fusion and fission energy applications, JOM (J. Occup. Med.) 66 (2014) 2458-2466.
  47. Y. Yamamoto, B. Pint, K. Terrani, K. Field, Y. Yang, L. Snead, Development and property evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors, J. Nucl. Mater. 467 (2015) 703-716. https://doi.org/10.1016/j.jnucmat.2015.10.019
  48. G.J. Youinou, R.S. Sen, Impact of accident-tolerant fuels and claddings on the overall fuel cycle: a preliminary systems analysis, Nucl. Technol. 188 (2014) 123-138. https://doi.org/10.13182/NT14-22
  49. C.H. Henager Jr., W.D. Bennett, A.L. Doherty, E. Fuller, J.S. Hardy, R.P. Omberg, Corrosion Report for the U-Mo Fuel Concept, Pacific Northwest National Laboratory (PNNL), Richland, WA (US), 2014.
  50. J. Hu, D. Wolfe, A. Motta, et al., Radiation Tolerance of Multilayer (TiN, TiAlN) Ceramic ATF Coating. 2017 ANS Winter Meeting, Washington, DC, United States, 2017.
  51. Y. Zhang, D.S. Aidhy, T. Varga, S. Moll, P.D. Edmondson, F. Namavar, K. Jin, C.N. Ostrouchov, W.J. Weber, The effect of electronic energy loss on irradiationinduced grain growth in nanocrystalline oxides, Phys. Chem. Chem. Phys. 16 (2014a) 8051-8059. https://doi.org/10.1039/C4CP00392F
  52. J. Brachet, C. Lorrette, A. Michaux, C. Sauder, I. Idarraga-Trujillo, M. Le Saux, A. Ambard, CEA studies on advanced nuclear fuel claddings for enhanced Accident Tolerant LWRs Fuel (LOCA and beyond LOCA conditions), in: Fontevraud 8, Contribution of Materials Investigations and Operating Experience to LWRs' Safety, Performance and Reliability, 2014. September 15-18, 2014, France, Avignon.
  53. K.A. Terrani, et al., Silicon carbide oxidation in steam up to 2 MPa, J. Am. Ceram. Soc. 97 (2014) 2331-2352. https://doi.org/10.1111/jace.13094
  54. Y. Katoh, et al., Radiation effects in SiC for nuclear structural applications, Curr. Opin. Solid State Mater. Sci. 16 (2012) 143-152. https://doi.org/10.1016/j.cossms.2012.03.005
  55. J.-H. Chun, S.-W. Lim, B.-D. Chung, W.-J. Lee, Safety evaluation of accidenttolerant FCM fueled core with SiC-coated zircalloy cladding for designbasis-accidents and beyond DBAs, Nucl. Eng. Des. 289 (2015) 287-295. https://doi.org/10.1016/j.nucengdes.2015.04.021
  56. T. Koyanagi, Y. Katoh, M. Snead, SiC/SiC Cladding Materials Properties Handbook, U.S. Department of Energy-Nuclear Technology Research and Development Advanced Fuels Campaign ORNL/TM-20, 2017.
  57. A. Ellison, J. Zhang, J. Peterson, A. Henry, Q. Wahab, J. Bergman, Y.N. Makarov, A. Vorob'ev, A. Vehanen, E. Janzen, High temperature CVD growth of SiC, Mater. Sci. Eng., B 61 (1999) 113-120. https://doi.org/10.1016/S0921-5107(98)00482-6
  58. H. Tsou, W. Kowbel, A hybrid PACVD SiC/CVD Si3N4SiC multilayer coating for oxidation protection of composites, Carbon 33 (1995) 1279-1288. https://doi.org/10.1016/0008-6223(95)00074-N
  59. Y. Katoh, K.A. Terrani, Systematic Technology Evaluation Program for SiC/SiC Composite Based Accident-Tolerant LWR Fuel Cladding and Core Structures: Revision 2015, Oak Ridge National Laboratory (ORNL), 2015.
  60. D.M. Carpenter, Assessment of Innovative Fuel Designs for High Performance Light Water Reactors, 2006.
  61. H. Feinroth, et al., Mechanical strength of CTP Triplex SiC fuel clad tubes after irradiation in MIT research reactor under PWR coolant conditions, Ceram. Eng. Sci. Proc. 30 (10) (2009) 47. https://doi.org/10.1002/9780470584002.ch4
  62. C. Deck, et al., Characterization of SiCeSiC composites for accident tolerant fuel cladding, J. Nucl. Mater. 466 (2015) 667-681. https://doi.org/10.1016/j.jnucmat.2015.08.020
  63. J.D. Stempien, et al., Characteristics of composite silicon carbide fuel cladding after irradiation under simulated PWR conditions, Nucl. Technol. 183 (2013) 13-29. https://doi.org/10.13182/NT12-86
  64. D. Kim, et al., Fabrication and measurement of hoop strength of SiC triplex tube for nuclear fuel cladding applications, J. Nucl. Mater. 458 (2015) 29-36. https://doi.org/10.1016/j.jnucmat.2014.11.117
  65. Y.I. Jung, S.-H. Kim, J.-Y. Park, Manufacturing process for the metal-ceramic hybrid fuel cladding tube, Trans. Korean Nucl. Soc. (2012) 25-26.
  66. M. Takeda, et al., Effect of hydrogen atmosphere on pyrolysis of cured polycarbosilane fibers, J. Am. Ceram. Soc. 83 (2000) 1063-1069. https://doi.org/10.1111/j.1151-2916.2000.tb01331.x
  67. H. Ichikawa, Development of high performance SiC fibers derived from polycarbosilane using electron beam irradiation curing-a review, J. Ceram. Soc. Jpn. 114 (2006) 455-460. https://doi.org/10.2109/jcersj.114.455
  68. T. Ishikawa, et al., High-strength alkali-resistant sintered SiC fibre stable to $2200^{\circ}C$, Nature 391 (1998) 773-775. https://doi.org/10.1038/35820
  69. Y. Katoh, et al., Continuous SiC fiber, CVI SiC matrix composites for nuclear applications: properties and irradiation effects, J. Nucl. Mater. 448 (2014) 448-476. https://doi.org/10.1016/j.jnucmat.2013.06.040
  70. C. Sauder, Ceramic matrix composites: nuclear applications, Ceram. Matrix Compos.: Mater. Model. Technol. (2014) 609-646.
  71. M. Uchihashi, et al., Development of SiC/SiC composites for nuclear reactor core with enhanced safety, in: International Conference on Nuclear Engineering (ICONE) 2015.23, The Japan Society of Mechanical Engineers, 2015.
  72. C.M. Parish, et al., Microstructure and hydrothermal corrosion behavior of NITE-SiC with various sintering additives in LWR coolant environments, J. Eur. Ceram. Soc. 37 (2017) 1261-1279. https://doi.org/10.1016/j.jeurceramsoc.2016.11.033
  73. J.Y. Park, et al., Long-term corrosion behavior of CVD SiC in 360 C water and $400^{\circ}C$ steam, J. Nucl. Mater. 443 (2013) 603-607. https://doi.org/10.1016/j.jnucmat.2013.07.058
  74. S. Kondo, M. Lee, T. Hinoki, Y. Hyodo, F. Kano, Effect of irradiation damage on hydrothermal corrosion of SiC, J. Nucl. Mater. 464 (2015) 36-42. https://doi.org/10.1016/j.jnucmat.2015.04.034
  75. Z. Duan, et al., Current status of materials development of nuclear fuel cladding tubes for light water reactors, Nucl. Eng. Des. 316 (2017) 131-150. https://doi.org/10.1016/j.nucengdes.2017.02.031
  76. I. Spitsberg, J. Steibel, Thermal and environmental barrier coatings for SiC/SiC CMCs in aircraft engine applications, Int. J. Appl. Ceram. Technol. (2004) 291-301.
  77. T. Maruyama, et al., Relationship between dimensional changes and the thermal conductivity of neutron irradiated SiC, J. Nucl. Mater. (2004) 329-333.
  78. L.L. Snead, Limits on irradiation-induced thermal conductivity and electrical resistivity in silicon carbide materials, J. Nucl. Mater. (2004) 524-529, 329-333.
  79. L.L. Snead, et al., Handbook of SiC properties for fuel performance modeling, J. Nucl. Mater. 371 (1) (2007) 329-377. https://doi.org/10.1016/j.jnucmat.2007.05.016
  80. R.H. Jones, D. Steiner, H.L. Heinisch, G.A. Newsome, H.M. Kerch, Radiation resistant ceramic matrix composites, J. Nucl. Mater. 245 (1997) 87-107. https://doi.org/10.1016/S0022-3115(97)00022-6
  81. K.A. Schwetz, in: Handbook of Ceramic Hard Materials, Wiley-VCH Verlag GmbH, 2008, pp. 683-748.
  82. N. Miriyala, et al., The mechanical behavior of a Nicalon/SiC composite at room temperature and $1000^{\circ}C$, J. Nucl. Mater. 253 (1998) 1-9. https://doi.org/10.1016/S0022-3115(97)00341-3
  83. E. Rohmer, E. Martin, C. Lorrette, Mechanical properties of SiC/SiC braided tubes for fuel cladding, J. Nucl. Mater. 453 (2014) 16-21. https://doi.org/10.1016/j.jnucmat.2014.06.035
  84. T. Koyanagi, Y. Katoh, Mechanical properties of SiC composites neutron irradiated under light water reactor relevant temperature and dose conditions, J. Nucl. Mater. 494 (2017) 46-54. https://doi.org/10.1016/j.jnucmat.2017.07.007
  85. Y. Katoh, et al., Thermophysical and mechanical properties of near-stoichiometric fiber CVI SiC/SiC composites after neutron irradiation at elevated temperatures, J. Nucl. Mater. 403 (2010) 48-61. https://doi.org/10.1016/j.jnucmat.2010.06.002
  86. G. Newsome, et al., Evaluation of neutron irradiated silicon carbide and silicon carbide composites, J. Nucl. Mater. 371 (2007) 76-89. https://doi.org/10.1016/j.jnucmat.2007.05.007
  87. T. Nozawa, et al., Determination and prediction of axial/off-axial mechanical properties of SiC/SiC composites, Fusion Eng. Des. 87 (2012) 803-807. https://doi.org/10.1016/j.fusengdes.2012.02.026
  88. C.K. Ang, et al., Examination of Hybrid Metal Coatings for Mitigation of Fission Product Release and Corrosion Protection of LWR SiC/SiC, Oak Ridge National Laboratory (ORNL), Oak Ridge, TN, United States, 2016.
  89. H.A. Friggens, D.R. Holmes, Nucleation and growth of magnetite films on Fe in high-temperature water, Corros. Sci. 8 (1968) 871-881. https://doi.org/10.1016/S0010-938X(68)80140-4
  90. C.S. Wukusick, J.F. Collins, An iron-chromium-aluminum alloy containing yttrium, Mater. Res. Std. 4 (1964).
  91. K.R. Robb, Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents, Oak Ridge National Lab.(ORNL), Oak Ridge, TN, United States, 2015.
  92. C.P. Massey, K.A. Terrani, S.N. Dryepondt, et al., Cladding burst behavior of Fe-based alloys under LOCA, J. Nucl. Mater. 470 (2016) 128-138. https://doi.org/10.1016/j.jnucmat.2015.12.018
  93. F.H. Stott, G.C. Wood, J. Stringer, The influence of alloying elements on the development and maintenance of protective scales, Oxid. Metals 44 (1995) 113-145. https://doi.org/10.1007/BF01046725
  94. J. Ejenstam, M. Thuvander, P. Olsson, et al., Microstructural stability of Fe-Cr-Al alloys at $450-550^{\circ}C$, J. Nucl. Mater. 457 (2015) 291-297. https://doi.org/10.1016/j.jnucmat.2014.11.101
  95. H.P. Qu, Y.P. Lang, C.F. Yao, et al., The effect of heat treatment on recrystallized microstructure, precipitation and ductility of hot-rolled Fe-Cr-Al-REM ferritic stainless steel sheets, Mater. Sci. Eng. 562 (2013) 9-16. https://doi.org/10.1016/j.msea.2012.11.008
  96. K.A. Gamble, T. Barani, D. Pizzocri, et al., An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions, J. Nucl. Mater. 491 (2017) 55-66. https://doi.org/10.1016/j.jnucmat.2017.04.039
  97. Kanthal APMT (tube) datasheet. http://kanthal.com/en/products/materialdatasheets/tube/kanthal-apmt/, 2012.
  98. B. Jonsson, Q. Lu, D. Chandrasekaran, et al., Oxidation and creep limited lifetime of Kanthal APMT(R), a dispersion strengthened FeCrAlMo alloy designed for strength and oxidation resistance at high temperatures, Oxid. Metals 79 (2013) 29-39. https://doi.org/10.1007/s11085-012-9324-4
  99. Z.T. Thompson, K.A. Terrani, Y. Yamamoto, ORNL/TM-2015/, Elastic Modulus Measurement of ORNL ATF FeCrAl Alloys, vol 632, 2015, pp. 1-17.
  100. S. Dryepondt, B.A. Pint, E. Lara-Curzio, Creep behavior of commercial FeCrAl foils: beneficial and detrimental effects of oxidation, Mater. Sci. Eng., A (2012) 10-18. https://doi.org/10.1016/S0921-5093(04)00900-1
  101. K.A. Terrani, B.A. Pint, Y.J. Kim, et al., Uniform corrosion of FeCrAl alloys in LWR coolant environments, J. Nucl. Mater. 479 (2016) 36-47. https://doi.org/10.1016/j.jnucmat.2016.06.047
  102. M.W. Barsoum, The $M_{N+1}AX_N$ phases: a new class of solids, Prog. Solid State Chem. 28 (2000) 201-281. https://doi.org/10.1016/S0079-6786(00)00006-6
  103. Z. Yanchun, S. Zhimei, S. Jihong, et al., Titanium silicon carbide: a ceramic or a metal, Z Metallkd 91 (2000) 329-334.
  104. P. Finkel, M.W. Barsoum, T. El-Raghy, Low temperature dependencies of the elastic properties of Ti 4 AlN 3, Ti 3 Al 1.1 C 1.8, and $Ti_3SiC_2$, J. Appl. Phys. 87 (4) (2000) 1701-1703. https://doi.org/10.1063/1.372080
  105. Z. Sun, Y. Zhou, M. Li, Oxidation behaviour of $Ti_3SiC_2$-based ceramic at $900-1300^{\circ}C$ in air, Corros. Sci. 43 (6) (2001) 1095-1109. https://doi.org/10.1016/S0010-938X(00)00142-6
  106. R. Radhakrishnan, J.J. Williams, M. Akinc, Synthesis and high-temperature stability of Ti3SiC2, J. Alloy. Comp. 285 (1-2) (1999) 85-88. https://doi.org/10.1016/S0925-8388(99)00003-1
  107. Z.M. Sun, H. Hashimoto, Z.F. Zhang, et al., Synthesis and characterization of a metallic ceramic material-Ti3SiC2, Mater. Trans. 47 (1) (2006) 170-174. https://doi.org/10.2320/matertrans.47.170
  108. T. El-Raghy, M.W. Barsoum, A. Zavaliangos, et al., Processing and mechanical properties of Ti3SiC2: II, effect of grain size and deformation temperature, J. Am. Ceram. Soc. 82 (10) (1999) 2855-2860. https://doi.org/10.1111/j.1151-2916.1999.tb02167.x
  109. Y. Zhou, Z. Sun, Microstructure and mechanism of damage tolerance for $Ti_3SiC_2$ bulk ceramics, Mater. Res. Innov. 2 (6) (1999) 360-363. https://doi.org/10.1007/s100190050114
  110. M.W. Barsoum, H.I. Yoo, I.K. Polushina, et al., Electrical conductivity, thermopower, and hall effect of $Ti_3AlC_2$, $Ti_4AlN_3$, and $Ti_3SiC_2$, Phys. Rev. B 62 (15) (2000) 10194. https://doi.org/10.1103/PhysRevB.62.10194
  111. A.S. Farle, C. Kwakernaak, S. van der Zwaag, W.G. Sloof, A conceptual study into the potential of $M_{n+1}AX_n$-phase ceramics for self-healing of crack damage, J. Eur. Ceram. Soc. 35 (2015).
  112. B. Li, Z. Yang, M. Chu, et al., $Ti_3SiC_2/UO_2$ composite pellets with superior high-temperature thermal conductivity, Ceram. Int. 44 (16) (2018) 19846-19850. https://doi.org/10.1016/j.ceramint.2018.07.244
  113. Y.W. Bao, Y.C. Zhou, et al., Mechanical properties of $Ti_3SiC_2$ at high temperature, Acta Metall. Sin. 17 (4) (2004) 465-470.
  114. M.W. Barsoum, T. El-Raghy, L.U.J.T. Ogbuji, Oxidation of Ti3SiC2 in air, J. Electrochem. Soc. 144 (7) (1997) 2508-2516. https://doi.org/10.1149/1.1837846
  115. H. Zhang, V. Presser, C. Berthold, et al., Mechanisms and kinetics of the hydrothermal oxidation of bulk titanium silicon carbide, J. Am. Ceram. Soc. 93 (4) (2010) 1148-1155. https://doi.org/10.1111/j.1551-2916.2009.03570.x
  116. W. Jiang, C.H. Henager Jr., T. Varga, et al., Diffusion of Ag, Au and Cs implants in MAX phase Ti3SiC2, J. Nucl. Mater. 462 (2015) 310-320. https://doi.org/10.1016/j.jnucmat.2015.04.002

Cited by

  1. Recent studies on potential accident-tolerant fuel-cladding systems in light water reactors vol.31, pp.3, 2020, https://doi.org/10.1007/s41365-020-0741-9
  2. Accident tolerant fuel thermal hydraulic behaviors evaluation during loss of coolant accident in CPR1000 vol.139, 2020, https://doi.org/10.1016/j.anucene.2019.107273
  3. Application and Development Progress of Cr-Based Surface Coatings in Nuclear Fuel Element: I. Selection, Preparation, and Characteristics of Coating Materials vol.10, pp.9, 2020, https://doi.org/10.3390/coatings10090808
  4. Three-Dimensional Modeling of Thermal-Mechanical Behavior of Accident Tolerant Fuels vol.9, 2020, https://doi.org/10.3389/fenrg.2021.636502
  5. Analysis of neutron physics and thermal hydraulics for fuel assembly of small modular reactor loaded with ATFs vol.152, 2021, https://doi.org/10.1016/j.anucene.2020.107957
  6. Structural, elastic, electronic, and anisotropic properties of Pbca-SiC and Pbcn-SiC vol.11, pp.4, 2021, https://doi.org/10.1063/5.0044672
  7. High-strength joint of nuclear-grade FeCrAl alloys achieved by friction stir welding and its strengthening mechanism vol.65, 2021, https://doi.org/10.1016/j.jmapro.2021.03.007
  8. Atomic-scale investigation of creep behavior and deformation mechanism in nanocrystalline FeCrAl alloys vol.206, 2020, https://doi.org/10.1016/j.matdes.2021.109766
  9. Silicon carbide coatings on zirconium alloy for light water reactor fuel cladding studies vol.1989, pp.1, 2020, https://doi.org/10.1088/1742-6596/1989/1/012009
  10. Thermal stability and performance of optimized ZrC x diffusion barriers in ceramic coating systems for ATF applications vol.104, pp.10, 2020, https://doi.org/10.1111/jace.17919
  11. Exploring the high-temperature steam oxidation behaviors of the lean-Cr (7-10 wt%) FeCrAl alloys vol.194, 2020, https://doi.org/10.1016/j.corsci.2021.109927
  12. Grid-to-rod fretting wear study of SiC/SiC composite accident-tolerant fuel claddings using an autoclave fretting bench test vol.488, 2020, https://doi.org/10.1016/j.wear.2021.204172