Abstract: Light water reactor (LWR) is the most widespreadly used type of nuclear power reactor, all of which are equipped with fuel claddings made of zirconium alloy. However, the Fukushima Daiichi accident has called into question the reliability of the in-service Zr alloy claddings, and thereby various international programs have been launched to develop accident tolerant fuel (ATF) and relevant candidate cladding materials. This paper summarizes the recent research progress of SiCf/SiC composites and Mo alloys cladding materials, along with the performance assessment with respect to LWR working conditions. In addition, the potential challenges for engineering practices and the future prospects are also discussed.
1 Wray P. Materials for nuclear energy in the post-Fukushima era: An interview with John Marra[J].American Ceramic Society Bulletin,2012,90:24. 2 Zinkle S J,Terrani K A, Gehin J C, et al. Accident tolerant fuels for LWRs: A perspective[J].Journal of Nuclear Materials,2014,448(1-3):374. 3 Yueh K, Carpene D, Feinroth H. Clad in clay[J].Nuclear Enginee-ring International,2010,55(666):14. 4 Terrani K A, Zinkle S J, Snead L L. Advanced oxidation-resistant iron-based alloys for LWR fuel cladding[J].Journal of Nuclear Materials,2014,448(1-3):420. 5 George N M, Terrani K A, Powers J J. Neutronic analysis of candidate accident-tolerant iron alloy cladding concepts[R].Tennessee: Oak Ridge National Laboratory,2013. 6 Pint B A, Terrani K A, Yamamoto Y, et al. Material selection for accident tolerant fuel cladding[J].Metallurgical and Materials Tran-sactions E,2015,2(3):190. 7 Younker I, Fratoni M. Neutronic evaluation of coating and cladding materials for accident tolerant fuels[J].Progress in Nuclear Energy,2016,88(1):10. 8 Deck C P, Jacobsen G M, Sheeder J, et al. Characterization of SiC-SiC composites for accident tolerant fuel cladding[J].Journal of Nuclear Materials,2015,466(1):667. 9 Katoh Y, Terrani K A. Systematic technology evaluation program for SiC/SiC composite-based accident-tolerant LWR fuel cladding and core structures: Revision 2015[R].Tennessee: Oka Ridge National Laboratory,2015. 10 Cheng B, Chou P, Kim Y J. Evaluations of Mo-alloy for light water reactor fuel cladding to enhance accident tolerance[J].Nuclear Sciences and Technologies,2016,2(5):1. 11 Cheng B, Kim Y J, Chou P. Improving accident tolerance of nuclear fuel with coated Mo-alloy cladding[J].Nuclear Engineering and Technology,2016,48(1):16. 12 Nelson A T, Sooby E S, Kim Y J, et al. High temperature oxidation of molybdenum in water vapor environments[J].Journal of Nuclear Materials,2014,448(1-3):441. 13 Duan Z G, Yang H L, Satoh Y, et al. Current status of materials development of nuclear fuel cladding tubes for light water reactors[J].Nuclear Engineering and Design,2017,316:131. 14 Kese K, Olsson P A T, Holston A M, et al. High temperature nanoindentation hardness and Young’s modulus measurement in a neutron-irradiated fuel cladding material[J].Journal of Nuclear Materials,2017,487(1):113. 15 Kim H G, Yang J H, Kim W J, et al. Development status of accident tolerant fuel for light water reactors in Korea[J].Nuclear Engineering and Technology,2016,48(1):1. 16 Ott L J, Robb K R, Wang D. Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions[J].Journal of Nuclear Mate-rials,2014,448(1-3):520. 17 Brown N R, Wysocki A, Terrani K, et al. The potential impact of enhanced accident tolerant cladding materials on reactivity initiated accidents in light water reactors[J].Annals of Nuclear Energy,2017,99(1):353. 18 Johnson K, Strom V, Wallenius J, et al. Oxidation of accident tole-rant fuel candidates[J].Journal of Nuclear Science and Technology,2017,54(3):280. 19 Lee Y H, Byun T S. A comparative study on the wear behaviors of cladding candidates for accident-tolerant fuel[J].Journal of Nuclear Materials,2015,465(1):857. 20 Sitten S B. Application of MELCOR to ATF concepts for sever accident analysis[R].San Antonio: GE Global Research Center,2014. 21 Katoh Y, Terrani K, Snead L L. Systematic technology evaluation program for SiC/SiC composite-based accident-tolerant LWR fuel cladding and core structures[C]∥Conference Record of the Accident Tolerant Fuel Concepts For Light Water Reactors. Tennessee,2014:277. 22 Silva C M, Katoh Y, Voit S,et al. Chemical reactivity of CVC and CVD SiC with UO2 at high temperatures[J].Journal of Nuclear Materials,2015,460:52. 23 Rohmer E, Martin E, Lorrette C. Mechanical properties of SiC/SiC braided tubes for fuel cladding[J].Journal of Nuclear Materials,2014,453(1-3):16. 24 Jacobson N, Harder B, Myers D. Oxidation transitions for SiC Part Ⅰ. Active-to-passive transitions[J].Journal of the American Ceramic Society,2013,96(3):838. 25 Harder B, Jacobson N, Myers D. Oxidation transitions for SiC Part Ⅱ. Passive-to-active transitions[J].Journal of the American Ceramic Society,2013,96(2):606. 26 Lee Y H, No H C, Lee J I. Design optimization of multi-layer silicon carbide cladding for light water reactors[J].Nuclear Engineering and Design,2017,311:213. 27 Kim D, Lee H J, Jang C H, et al. Influence of microstructure on hydrothermal corrosion of chemically vapor processed SiC composite tubes[J].Journal of Nuclear Materials,2017,492(1):6. 28 Charpentier L, Dawi K, Balat-Pichelin M, et al. Chemical degradation of SiC/SiC composite for the cladding of gas-cooled fast reactor in case of severe accident scenarios[J].Corrosion Science,2012,59(3):127. 29 Braun J, Gueneau C, Alpettaz T, et al. Chemical compatibility between UO2 fuel and SiC cladding for LWRs application to ATF (accident-tolerant fuels)[J].Journal of Nuclear Materials,2017,487(1):380. 30 Matsumiya H, Yoshioka K, Kikuchi T, et al. Reactivity measurements of SiC for accident-tolerant fuel[J].Progress in Nuclear Energy,2015,82(1):16. 31 Yueh K, Terrani K A. Silicon carbide composite for light water reactor fuel assembly applications[J].Journal of Nuclear Materials,2014,448(1-3):380. 32 Kim W J, Kim D, Park J Y. Fabrication and material issues for the application of SiC composites to LWR fuel cladding[J].Nuclear Engineering and Technology,2013,45(4):565. 33 Koyanagi T, Katoh Y, Terrani K A, et al. Hydrothermal corrosion of silicon carbide joints without radiation[J].Journal of Nuclear Materials,2016,481(1):226. 34 Gamble K A, Barani T, Pizzocri D, et al. An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions[J].Journal of Nuclear Materials,2017,491(1):55. 35 Wu X, Kozlowski T, Hales J D. Neutronics and fuel performance evaluation of accident tolerant FeCrAl cladding under normal operation conditions[J].Annals of Nuclear Energy,2015,85(1):763. 36 Guria A, Charit I. Tensile properties of accident-tolerant aluminum-bearing ferritic steels[J].Annals of Nuclear Energy,2017,100(P1):82. 37 Yano Y, Tanno T, Oka H, et al. Ultra-high temperature tensile properties of ODS steel claddings under severe accident conditions[J].Journal of Nuclear Materials,2017,487(1):229. 38 Chun J H, Lim S W, Chung B D, et al. Safety evaluation of accident-tolerant FCM fueled core with SiC-coated zircalloy cladding for design-basis-accidents and beyond DBAs[J].Nuclear Engineering and Design,2015,289:287. 39 Wang X, Zheng M J, Szlufarska I, et al. Continuum model for hydrogen pickup in zirconium alloys of LWR fuel cladding[J].Journal of Applied Physics,2017,121(13):135101. 40 Katoh Y, Snead L L, Szlufarska I, et al. Radiation effects in SiC for nuclear structural applications[J].Current Opinion in Solid State and Materials Science,2012,16(3):143. 41 Sitton S B, Hurley D, Khafizov M, et al. Silicon carbide gap analysis and feasibility study[R].Idaho: Idaho National Laboratory,2013. 42 Stone J G, Schleicher R, Deck C P, et al. Stress analysis and probabilistic assessment of multi-layer SiC-based accident tolerant nuclear fuel cladding[J].Journal of Nuclear Materials,2015,466(1):682. 43 Stempien J D, Carpenter D M, Kohse G, et al. Characteristics of composite silicon carbide fuel cladding after irradiation under simulated PWR conditions[J].Nuclear Technology,2012,183(1):13. 44 Terrani K A, Yang Y, Kim Y J, et al. Hydrothermal corrosion of SiC in LWR coolant environments in the absence of irradiation[J].Journal of Nuclear Materials,2015,465(1):488. 45 Park J Y, Kim I H, Jung Y I, et al. Long-term corrosion behavior of CVD SiC in 360 ℃ water and 400 ℃ steam[J].Journal of Nuclear Materials.2013,443(1-3):603. 46 Katoh Y, Ozawa K, Shih C, et al. Continuous SiC fiber, CVI SiC matrix composites for Nuclear applications: Properties and irradiation Effects[J].Journal of Nuclear Materials,2014,448(1-3):448. 47 Belgacem M B, Richet V, Terrani K A, et al. Thermo-mechanical analysis of LWR SiC/SiC composite cladding[J].Journal of Nuclear Materials,2014,447(1-3):125. 48 El-Genk M S, Tournier J M. A review of refractory metal alloys and mechanically alloyed-oxide dispersion strengthened steels for space nuclear power systems[J].Journal of Nuclear Materials,2005,340(1):93. 49 Cheng P M, Zhang G J, Zhang J Y, et al. Coupling effect of intergranular and intragranular particles on ductile fracture of Mo-La2O3 alloys[J].Materials Science and Engineering A,2015,640(1):320. 50 Cockeram B V. The fracture toughness and toughening mechanism of commercially available unalloyed molybdenum and oxide dispersion strengthened molybdenum with an equiaxed, large grain structure[J].Metallurgical and Materials Transaction A,2009,40(12):2843. 51 Cockeram B V, Smith R W, Hashimoto N, et al. The swelling, microstructure, and hardening of wrought LCAC, TZM, and ODS molybdenum following neutron irradiation[J].Journal of Nuclear Materials,2011,418(1-3):121. 52 Byun T S, Li M, Cockeram B V, et al. Deformation and fracture properties in neutron irradiated pure Mo and Mo alloys[J].Journal of Nuclear Materials,2008,376(2):240.