MATERIALS AND SUSTAINABLE DEVELOPMENT: ADVANCED MATERIALS FOR CLEAN ENERGY UTILIZATION |
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A Complete Review and a Prospect on the Candidate Materials for Accident-tolerant Fuel Claddings |
LIU Junkai1, ZHANG Xinhu1, YUN Di1,2
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1 School of Energy and Power Engineering, Xi’an Jiaotong University, Xi’an 710049; 2 State Key Laboratory of Multi-phase Flow in Power Engineering, Xi’an Jiaotong University, Xi’an 710049 |
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Abstract In the 2011 Fukushima nuclear power plant accident, the zirconium alloy claddings in the reactor fuel were oxidized violently by high temperature water vapor generated under accident conditions, resulting in a large amount of hydrogen and heat, which eventually led to reactor core melt down and hydrogen explosion. The event delivered severe impacts on the human society and the environment. From then on, the development programs of accident-tolerant fuels (ATF) have been initiated worldwide. Compared with the traditional UO2-Zr system, ATF can sustain or improve the fuel performance at normal operating conditions, as well as maintain the core integrity for an extended time after accidents to provide temporal safety margins for the accident management.
The environment of the reactor core is very harsh, as the claddings are exposed lastingly to high temperature and high pressure corrosion medium, and moreover, neutron irradiation. As a result, if any new materials were to qualify as cladding materials, they would need excellent radiation stability and corrosion resistance. According to the research and evaluation of different researchers, the accident-tolerant fuel cladding materials which has displayed the potential of substituting for Zr alloy can be classified into ceramic materials and metallic materials, in which the ceramic materials are mainly represented by SiC/SiC composite materials, and the metallic materials mainly include Fe-based alloy exemplified by FeCrAl and refractory metals exemplified by Mo/Mo alloy.
Each of the three materials has its own advantages and disadvantages, and still needs improvements to attain the requirement of engineering application, and all of them have some unresolved key issues. The research and development of the FeCrAl alloy has reached the second generation model alloy. The thermophysical and mechanical properties, corrosion resistance and radiation resis-tance of FeCrAl alloy are outstanding, while the research on hydrogen permeability and industrial processing and welding are still in progress. In the case of SiC/SiC composites, the high brittleness of SiC material leads to inadequate mechanical strength. Hence researchers have proposed various structural design schemes to reduce the failure probability of cladding tubes, but the final structural design of the cladding remains undetermined. The sharp reduction of thermal conductivity of SiC/SiC composites caused by irradiation and the joining and fabrication are still under investigation. Mo and Mo alloys possesses excellent mechanical properties and radiation resistance, but exhibits poor corrosion resistance. The present countermeasures are mainly focused on improving the purity of molybdenum, adjusting the elemental composition of the alloys and adopting surface coating techniques. The processability of these three kinds of cladding tube materials has not reached the level of industrial manufacturing of thin-wall long tubes. For these candidate materials, a property database and an integrated series of standards are of great necessity to evaluate the quality of the materials. In addition, the requirement of in-core behavior assessment also necessitates the development of the corresponding fuel performance code.
This paper summarizes the latest research progress on the candidate ATF cladding materials, including their physical properties, corrosion resistance, mechanical behaviors, radiation resistance, pellet-cladding mechanical and chemical interactions, behaviors under accident conditions and engineering applications. It also gives critical discussions about the current research situation and the potential key issues of each candidate material.
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Published: 20 July 2018
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1 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. 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 Sowder A G. Challenges and opportunities for commercialization of enhanced accident tolerant fuel for light water reactors: A utility-informed perspective[J].IAEA TECDOC Series,119. 4 Cheng B, Chou P, Kim Y J. Development of Mo-based accident to-lerant LWR fuel cladding[J].IAEA TECDOC Series,2016:66. 5 Ejenstam J, Halvarsson M, Weidow J, et al. Oxidation studies of Fe10CrAl-RE alloys exposed to Pb at 550 ℃ for 10 000 h[J].Journal of Nuclear Materials,2013,443(1-3):161. 6 Yamamoto Y, Field K G, Snead L L. Optimization of Nuclear Grade FeCrAl fuel cladding for light water reactors[J].IAEA TECDOC Series,2016:55. 7 Yamamoto Y, Pint B A, Terrani K A, et al. Development and pro-perty evaluation of nuclear grade wrought FeCrAl fuel cladding for light water reactors[J].Journal of Nuclear Materials,2015,467:703. 8 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. 9 George N M, Terrani K, Powers J, et al. Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors[J].Annals of Nuclear Energy,2015,75:703. 10 Hu X, Terrani K A, Wirth B D, et al. Hydrogen permeation in FeCrAl alloys for LWR cladding application[J].Journal of Nuclear Materials,2015,461:282. 11 Field K G, Hu X, Littrell K, et al. Stability of Model Fe-Cr-Al alloys under the presence of neutron radiation [R].Oak Ridge National Lab. (ORNL), Oak Ridge, TN, USA,2014. 12 Field K G, Hu X, Littrell K C, et al. Radiation tolerance of neutron-irradiated model Fe-Cr-Al alloys[J].Journal of Nuclear Materials,2015,465:746. 13 Pint B A. Optimization of reactive-element additions to improve oxidation performance of alumina-forming alloys[J].Journal of the American Ceramic Society,2003,86(4):686. 14 Braase L, May W E. Advanced fuels campaign FY 2014 accomplishments report[R].Idaho National Lab. (INL), Idaho Falls, ID, USA,2014. 15 Dryepondt S N, Hoelzer D T, Pint B A, et al. Development of ODS FeCrAl alloys for accident-tolerant fuel cladding[R].Oak Ridge National Lab. (ORNL), Oak Ridge, TN, USA,2015. 16 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,2017,491:55. 17 Powers J J, Worrall A, Robb K R, et al. ORNL analysis of operational and safety performance for candidate accident tolerant fuel and cladding concepts[J].IAEA TECDOC Series,2016:253. 18 Yamamoto Y, Gussev M N, Kim B, et al. Optimized properties on base metal and thin-walled tube of Generation II ATF FeCrAl[R].Oak Ridge National Lab.(ORNL), Oak Ridge, TN, USA,2015. 19 潘钱付,刘超红.国内院企合作首次研制出耐事故燃料全尺寸FeCrAl包壳管[EB/OL].[2018-04-15]http://mp.weixin.qq.com/s/DNAonyVd7d4prUCfH7q4nw. 20 Mcmurray J W, Hu R, Ushakov S V, et al. Solid-liquid phase equilibria of Fe-Cr-Al alloys and spinels[J].Journal of Nuclear Materials,2017,492:128. 21 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,2015,85:763. 22 Terrani K A, Karlsen T M, Yamamoto Y. Input correlations for irradiation creep of FeCrAl and SiC based on in-pile Halden test results[R].Oak Ridge National Lab. (ORNL), Oak Ridge, TN, USA,2016. 23 Saunders S, Evans H E, Li M, et al. Oxidation growth stresses in an alumina-forming ferritic steel measured by creep deflection[J].Oxidation of Metals,1997,48(3-4):189,. 24 Massey C P, Terrani K A, Dryepondt S N, et al. Cladding burst behavior of Fe-based alloys under LOCA[J].Journal of Nuclear Mate-rials,2016,470:128. 25 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,2017,487:229. 26 Gussev M N, Byun T S, Yamamoto Y, et al. In-situ tube burst testing and high-temperature deformation behavior of candidate mate-rials for accident tolerant fuel cladding[J].Journal of Nuclear Mate-rials,2015,466:417. 27 Rybicki G C, Smialek J L. Effect of the θ-α-Al2O3 transformation on the oxidation behavior of β-NiAl+ Zr[J].Oxidation of Metals,1989,31(3):275. 28 Opila E J, Myers D L. Alumina volatility in water vapor at elevated temperatures[J].Journal of the American Ceramic Society,2004,87(9):1701. 29 Pint B A, Terrani K A, Brady M P, et al. High temperature oxidation of fuel cladding candidate materials in steam-hydrogen environments[J].Journal of Nuclear Materials,2013,440(1-3):420. 30 Terrani K A, Pint B A, Kim Y J, et al. Uniform corrosion of FeCrAl alloys in LWR coolant environments[J].Journal of Nuclear Materials,2016,479:36. 31 Pint B A, Terrani K A, Brady M P, et al. High temperature oxidation of fuel cladding candidate materials in steam-hydrogen environments[J].Journal of Nuclear Materials,2013,440(1-3):420. 32 Pan D, Zhang R, Wang H, et al. Formation and stability of oxide layer in FeCrAl fuel cladding material under high-temperature steam[J].Journal of Alloys and Compounds,2016,684:549. 33 Badini C, Laurella F. Oxidation of FeCrAl alloy: Influence of temperature and atmosphere on scale growth rate and mechanism[J].Surface and Coatings Technology,2001,135(2):291. 34 Xu Y, Zhao S, Liu F, et al. Studies on oxidation and deuterium permeation behavior of a low temperature α-Al2O3-forming FeCrAl ferritic steel[J].Journal of Nuclear Materials,2016,477:257. 35 Sooby Wood E, Terrani K A, Nelson A T. Sensitivity of measured steam oxidation kinetics to atmospheric control and impurities[J].Journal of Nuclear Materials,2016,477:228. 36 Field K G, Briggs S A, Sridharan K, et al. Mechanical properties of neutron-irradiated model and commercial FeCrAl alloys[J].Journal of Nuclear Materials,2017,489:118. 37 Field K G, Gussev M N, Yamamoto Y, et al. Deformation behavior of laser welds in high temperature oxidation resistant Fe-Cr-Al alloys for fuel cladding applications[J].Journal of Nuclear Materials,2014,454(1-3):352. 38 Regina J R, Dupont J N, Marder A R. The effect of chromium on the weldability and microstructure of Fe-Cr-Al weld cladding[J].Welding Journal,2007,86(6):170. 39 Braase L A, Carmack W J. Advanced Fuels Campaign FY 2015 Accomplishments Report[R].Idaho National Lab. (INL), Idaho Falls, ID, USA,2015. 40 Katoh Y, Snead L L, Henager C H, et al. Current status and recent research achievements in SiC/SiC composites[J].Journal of Nuclear Materials,2014,455(1-3):387. 41 Snead L L, Nozawa T, Katoh Y, et al. Handbook of SiC properties for fuel performance modeling[J].Journal of Nuclear Materials,2007,371(1-3):329. 42 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. 43 Lee Y, 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. 44 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. 45 Petrie C M, Koyanagi T, Mcduffee J L, et al. Experimental design and analysis for irradiation of SiC/SiC composite tubes under a prototypic high heat flux[J].Journal of Nuclear Materials,2017,491:94. 46 Koyanagi T, Katoh Y. Mechanical properties of SiC composites neutron irradiated under light water reactor relevant temperature and dose conditions[J].Journal of Nuclear Materials,2017,494:46. 47 Koyanagi T, Ozawa K, Hinoki T, et al. Effects of neutron irradiation on mechanical properties of silicon carbide composites fabricated by nano-infiltration and transient eutectic-phase process[J].Journal of Nuclear Materials,2014,448(1-3):478. 48 Mazzoccoli J P, Choi J, Xu P. Progress on the Westinghouse Accident Tolerant Fuel Programme[J].IAEA TECDOC Series,2016:286. 49 Lee Y, Kazimi M S. A structural model for multi-layered ceramic cylinders and its application to silicon carbide cladding of light water reactor fuel[J].Journal of Nuclear Materials,2015,458:87. 50 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,2013,183(1):13. 51 Li B. Pellet cladding mechanical interactions of ceramic claddings fuels under light water reactor conditions[D].Columbia, South Carolina, USA: University of South Carolina,2013. 52 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].Oak Ridge National Laboratory. (ORNL), Oak Ridge,TN,USA,2015. 53 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,2015,466:667. 54 Kohyama A, Kishimoto H, Park J S, et al. Irradiation programme of SiC/SiC fuel claddings at Halden reactor[J].IAEA TECDOC Series,2016:302. 55 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,2015,466:682. 56 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. 57 Katoh Y, Dong S M, Kohyama A. Thermo-mechanical properties and microstructure of silicon carbide composites fabricated by nano-infiltrated transient eutectoid process[J].Fusion Engineering and Design,2002,61-62(Supplement C):723. 58 Katoh Y, Nozawa T, Snead L L, et al. Stability of SiC and its composites at high neutron fluence[J].Journal of Nuclear Materials,2011,417(1-3):400. 59 Jacobsen G M, Stone J D, Khalifa H E, et al. Investigation of the C-ring test for measuring hoop tensile strength of nuclear grade ceramic composites[J].Journal of Nuclear Materials,2014,452(1-3):125. 60 Ben-Belgacem M, 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. 61 Lee Y. Safety of light water reactor fuel with silicon carbide cladding[D].Cambridge, Massachusetts, USA: Massachusetts Institute of Technology,2013. 62 Alva L, Shapovalov K, Jacobsen G M, et al. Experimental study of thermo-mechanical behavior of SiC composite tubing under high temperature gradient using solid surrogate[J].Journal of Nuclear Mate-rials,2015,466:698. 63 Angelici Avincola V, Guenoun P, Shirvan K. Mechanical perfor-mance of SiC three-layer cladding in PWRs[J].Nuclear Engineering and Design,2016,310:280. 64 Lee Y, Mckrell T J, Kazimi M S. Thermal shock fracture of hot silicon carbide immersed in water[J].Journal of Nuclear Materials,2015,467:172. 65 Park J, Kim I, Jung Y, et al. Long-term corrosion behavior of CVD SiC in 360 ℃ water and 400 ℃ steam[J].Journal of Nuclear Mate-rials,2013,443(1-3):603. 66 Henager C H, Schemer-Kohrn A L, Pitman S G, et al. Pitting corrosion in CVD SiC at 300 ℃ in deoxygenated high-purity water[J].Journal of Nuclear Materials,2008,378(1):9. 67 Kim D, Lee H J, Jang C, et al. Influence of microstructure on hydrothermal corrosion of chemically vapor processed SiC composite tubes[J].Journal of Nuclear Materials,2017,492:6. 68 Kim W, Hwang H S, Park J Y, et al. Corrosion behaviors of sintered and chemically vapor deposited silicon carbide ceramics in water at 360 ℃[J].Journal of Materials Science Letters,2003,22(8):581. 69 Kim W, Hwang H S, Park J Y. Corrosion behavior of reaction-bonded silicon carbide ceramics in high-temperature water[J].Journal of Materials Science Letters,2002,21(9):733. 70 Parish C M, Terrani K A, Kim Y, et al. Microstructure and hydrothermal corrosion behavior of NITE-SiC with various sintering additives in LWR coolant environments[J].Journal of the European Ceramic Society,2017,37(4):1261. 71 Terrani K A, Pint B A, Parish C M, et al. Silicon carbide oxidation in steam up to 2 MPa[J].Journal of the American Ceramic Society,2014,97(8):2331. 72 Kondo S, Lee M H, Hinoki T. Acceleration of the hot water corrosion of SiC by Ion 7.7. irradiation[J].IAEA TECDOC Series,2016:329. 73 Opila E J. Oxidation and volatilization of silica formers in water vapor[J].Journal of the American Ceramic Society,2003,86(8):1238. 74 Lee Y, Mckrell T J, Yue C, et al. Safety assessment of SiC cladding oxidation under loss-of-coolant accident conditions in light water reactors[J].Nuclear Technology,2013,183(2):210. 75 Hinoki T, Lee M H, Kano F, et al. Effect of constituents of silicon carbide composites on oxidation behaviour[J].IAEA TECDOC Series,2016:314. 76 Snead L L, Nozawa T, Katoh Y, et al. Handbook of SiC properties for fuel performance modeling[J].Journal of Nuclear Materials,2007,371(1-3):329. 77 Chun J, Lim S, Chung B, et al. Safety evaluation of accident-tole-rant FCM fueled core with SiC-coated zircalloy cladding for design-basis-accidents and beyond DBAs[J].Nuclear Engineering and Design,2015,289:287. 78 Snead L L, Zinkle S J, White D P. Thermal conductivity degradation of ceramic materials due to low temperature, low dose neutron irradiation[J].Journal of Nuclear Materials,2005,340(2-3):187. 79 Ahn K. Comparison of silicon carbide and zircaloy4 cladding during LBLOCA[D].Cambridge, Massachusetts, USA: Massachusetts Institute of Technology,2006,4. 80 Katoh Y, Hashimoto N, Kondo S, et al. Microstructural development in cubic silicon carbide during irradiation at elevated temperatures[J].Journal of Nuclear Materials,2006,351(1-3):228. 81 Snead L L, Katoh Y, Connery S. Swelling of SiC at intermediate and high irradiation temperatures[J].Journal of Nuclear Materials,2007,367-370:677. 82 Kondo S, Katoh Y, Snead L L. Microstructural defects in SiC neutron irradiated at very high temperatures[J].Journal of Nuclear Materials,2008,382(2-3):160. 83 Katoh Y, Ozawa K, Hinoki T, et al. Mechanical properties of advanced SiC fiber composites irradiated at very high temperatures[J].Journal of Nuclear Materials,2011,417(1-3):416. 84 Braun J, Guéneau 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,2017,487:380. 85 Silva C M, Katoh Y, Voit S L, et al. Chemical reactivity of CVC and CVD SiC with UO2 at high temperatures[J].Journal of Nuclear Materials,2015,460:52. 86 Lee Y, Kim H S, No H C. Failure probabilities of SiC clad fuel du-ring a LOCA in public acceptable simple SMR (PASS)[J].Nuclear Engineering and Design,2015,292:1. 87 Deng Y, Wu Y, Qiu B, et al. Development of a new pellet-clad mechanical interaction (PCMI) model and its application in ATFs[J].Annals of Nuclear Energy,2017,104:146. 88 Katoh Y, Snead L L, Cheng T, et al. Radiation-tolerant joining technologies for silicon carbide ceramics and composites[J].Journal of Nuclear Materials,2014,448(1-3):497. 89 Ferraris M, Salvo M, Casalegno V, et al. Joining of SiC-based materials for nuclear energy applications[J].Journal of Nuclear Materials,2011,417(1-3):379. 90 Koyanagi T, Katoh Y, Terrani K A, et al. Hydrothermal corrosion of silicon carbide joints without radiation[J].Journal of Nuclear Materials,2016,481:226. 91 Koyanagi T, Katoh Y, Kiggans J O, et al. Irradiation resistance of silicon carbide joint at light water reactor-relevant temperature[J].Journal of Nuclear Materials,2017,488:150. 92 Khalifa H E, Koyanagi T, Jacobsen G M, et al. Radiation stable, hybrid, chemical vapor infiltration/preceramic polymer joining of si-licon carbide components[J].Journal of Nuclear Materials,2017,487:91. 93 Causey R A, Wampler W R, Retelle J R, et al. Tritium migration in vapor-deposited β-silicon carbide[J].Journal of nuclear materials,1993,203(3):196. 94 Jung P. Diffusion and retention of helium in graphite and silicon carbide[J].Journal of Nuclear Materials,1992,191:377. 95 Farmer M T, Leibowitz L, Terrani K A, et al. Scoping assessments of ATF impact on late-stage accident progression including molten core-concrete interaction[J].Journal of Nuclear Materials,2014,448(1-3):534. 96 Farmer M T, Leibowitz L, Terrani K A, et al. Scoping assessments of ATF impact on late-stage accident progression including molten core-concrete interaction[J].Journal of Nuclear Materials,2014,448(1-3):534. 97 Brown N R, Wysocki A J, Terrani K A, et al. The potential impact of enhanced accident tolerant cladding materials on reactivity initiated accidents in light water reactors[J].Annals of Nuclear Energy,2017,99:353. 98 Cockeram B V, Smith R W, Leonard K J, et al. Irradiation harde-ning in unalloyed and ODS molybdenum during low dose neutron irradiation at 300 ℃ and 600 ℃[J].Journal of Nuclear Materials,2008,382(1):1. 99 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. 100 Cheng B, Chou P, Kim Y. Evaluations of Mo-alloy for light water reactor fuel cladding to enhance accident tolerance[J].EPJ Nuclear Sciences & Technologies,2016,2:5. 101 Cheng B, Kim Y, Chou P. Improving accident tolerance of nuclear fuel with coated Mo-alloy cladding[J].Nuclear Engineering and Technology,2016,48(1):16. 102 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. |
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