Selection of fuel cladding material for nuclear fission reactors

Engineering Failure Analysis

Volumn 18, Issue 8, 2011, Pages 1943-1962

  Azevedo, C R F ^a  

^a UNIVERSITY OF SÃO PAULO   (Brazil)

Author keywords

Failure modes; Fuel cladding; Materials selection; Nuclear fission reactor

Indexed keywords

ABSORPTION CROSS SECTIONS; CARBON EMISSIONS; CHEMICAL COMPATIBILITY; DESIGN CONSTRAINTS; FUEL CLADDING; FUEL CYCLE; GENERATION IV REACTORS; MATERIALS SELECTION; NEUTRON RADIATIONS; NUCLEAR FISSION REACTOR; NUCLEAR WASTE; RADIOACTIVE FISSION PRODUCTS; SAFETY BARRIERS; SERVICE TEMPERATURE;

CARBON FIBER REINFORCED PLASTICS; ENERGY CONVERSION; ENERGY EFFICIENCY; FAILURE ANALYSIS; FISSION PRODUCTS; FUELS; GLOBAL WARMING; NUCLEAR ENERGY; NUCLEAR FUEL ELEMENTS; NUCLEAR PROPULSION; RADIOACTIVE WASTES; REACTOR SHUTDOWNS; THERMAL CONDUCTIVITY; THERMAL EXPANSION;

CLADDING (COATING);

EID: 80052578363     PISSN: 13506307     EISSN: None     Source Type: Journal    
DOI: 10.1016/j.engfailanal.2011.06.010     Document Type: Review

References (107)

1. Ashby MF, Smidmana M. Materials for nuclear power systems. Granta Design White Paper; 2011. http://teaching.grantadesign.com//download/Teachingmaterials/Papers-nuclear.pdf.
2. Renewable 2007: Global status report. [accessed 19.07.10]. http://www.ren21.net/pdf.
3. Life-cycle assessment. Vattenfall's electricity in Sweden. [accessed 19.07.10]. http://www.vattenfall.com.
4. Nuclear technology review 2009. [accessed 19.07.10]. http://www.iaea.org/Publications/Reports/ntr2009.pdf.
5. Jacoby M. Coming back to nuclear energy. Chem Eng News 2009, 87(34):14-18. [accessed 19.07.10].
6. Chater J. A history of nuclear power. [accessed 19.07.10]. http://www.focus-nuclear.com/pdf/TP_History_nuclear.pdf.
7. Survey of national programs for managing high-level radioactive waste and spent nuclear fuel. [accessed 19.07.10]. http://www.nwtrb.gov/reports/nwtrb%20sept%2009.pdf.
8. Lake J.A., Bennett R.G., Kotek J.F. Next generation nuclear power. Sci Am 2002, 286(1):72-79.
9. Howe J.P. The beginning of nuclear materials: studies of corrosion and cladding. J Nucl Mater 1981, 100:36-48.
10. Smith C.S. Some recollections of metallurgy at Los Alamos, 1943-45. J Nucl Mater 1981, 100:3-10.
11. Simnad M.T. A brief history of power reactor fuels. J Nucl Mater 1981, 100:93-107.
12. Burke J.E. Recollections of processing uranium hydride and plutonium at wartime Los Alamos. J Nucl Mater 1981, 100:93-107.
13. Lustman B. Development of the Zircaloy-clad UO2 fuel element for Shippingport. J Nucl Mater 1981, 100:72-77.
14. Outline history of nuclear energy. Nuclear world association. [accessed 19.07.10]. http://www.world-nuclear.org/info/inf54.html.
15. Nuclear power in the United Kingdom. [accessed 19.07.10]. http://www.world-nuclear.org/info/default.aspx?id=382&terms=united+kingdom.
16. A white paper on nuclear power. [accessed 19.07.10]. http://www.berr.gov.uk/files/file43006.pdf.
17. US nuclear power policy. [accessed 19.07.10]. http://www.world-nuclear.org/info/default.aspx?id=25741&terms=united+states.
18. Nuclear power in China. [accessed 19.07.10]. http://www.world-nuclear.org/info/default.aspx?id=320&terms=china.
19. Nuclear power in India. [accessed 19.07.10]. http://www.world-nuclear.org/info/default.aspx?id=338&terms=india.
20. Nuclear technology review. [accessed 19.07.10]. http://www.iaea.org/Publications/Reports/ntr2009.pdf.
21. Next-generation nuclear energy: the ESBWR. [accessed 19.07.10]. http://www.ans.org/pubs/magazines/nn/docs/2006-1-3.pdf.
22. Olander D.R. Nuclear fuels - present and future. J Nucl Mater 2009, 389:1-22.
23. Oka Y., Koshizuka S. Conceptual design study of advanced power reactors. Prog Nucl Energy 1998, 32:163-177.
24. Abu-Khader M.M. Recent advances in nuclear power: a review. Prog Nucl Energy 2009, 51:225-235.
25. Rouben B. Introduction to reactor physics. [accessed 19.07.10]. http://canteach.candu.org/library/20040501.pdf.
26. DOE Fundamentals handbook. Nuclear physics and reactor theory, vol. 1. [accessed 19.07.10]. http://www.hss.energy.gov/nuclearsafety/ns/techstds/standard/hdbk1019/h1019v1.pdf.
27. DOE Fundamentals handbook. Nuclear physics and reactor theory, vol. 2. [accessed 19.07.10]. http://www.hss.energy.gov/nuclearsafety/ns/techstds/standard/hdbk1019/h1019v2.pdf.
28. Materials for nuclear power generation. [accessed 19.07.10]. http://www.doitpoms.ac.uk/tlplib/nuclear_materials/index.php.
29. Development of radiation resistant reactor core structural materials. [accessed 19.07.10]. http://www.iaea.org/About/Policy/GC/GC51/GC51InfDocuments/English/gc51inf-3-att7_en.pdf.
30. Robertson J.A.L. Zirconium: an international nuclear material. J Nucl Mater 1981, 100:108-118.
31. Northwood D.O. The development and application of zirconium alloys. Mater Des 1985, 6(2):58-70.
32. Motta A.T., Yilmazbayhan A., Silva M.J.G., Comstock R.J., Was G.S., Busby J.T., et al. Zirconium alloys for supercritical water reactor applications: challenges and possibilities. J Nucl Mater 2007, 371:61-75.
33. Wikmark G., Hallstadius L., Yueh K. Cladding to sustain corrosion, creep and growth at high burn-ups. Nucl Eng Technol 2009, 41(2):143-148.
34. Cambridge Engineering Selector. CES Edupack software; 2010.
35. Azevedo CRF. A review on neutron-irradiation-induced hardening of metallic components, Eng Fail Anal 2011, doi:10.1016/j.engfailanal.2011.06.008. doi:10.1016/j.engfailanal.2011.06.008.
36. Ells C.E., Perryman E.C.W. Effects of neutron-induced gas formation on beryllium. J Nucl Mater 1959, 1:73-84.
37. Raine T., Robinson J.A. The development and properties of beryllium-calcium alloys for use in high temperature, high pressure carbon dioxide. J Nucl Mater 1962, 7(3):263-278.
38. 2 by autoradiographic techniques. J Nucl Mater 1965, 15(4):249-262.
39. Wanklyn J.N., Jones P.J. The aqueous corrosion of reactor materials. J Nucl Mater 1962, 6:291-329.
40. Snead L.L., Zinkle S.J. Use of beryllium and beryllium oxide in space reactors. AIP Conf Proc 2005, 746(February 6):768-775. Space technology and applications int. forum-staif 2005: Conf. Thermophys in Micrograv; Conf Comm/Civil Next Gen. Space Transp; 22nd Symp Space Nucl. Powr Propuls.; Conf.Human/Robotic Techn. Nat'l Vision Space Expl.; 3rd Symp Space Colon.; 2nd Symp. New Frontiers; doi:10.1063/1.1867196.
41. US Department of Energy. The Manhattan Project. An interactive history. [accessed 19.07.10]. http://www.cfo.doe.gov/me70/manhattan/final_reactor_x-10.htm.
42. Simnad M.T. Perspective on research reactors and their fuel elements. J Nucl Mater 1981, 100:78-92.
43. Eatherly W.P. Nuclear graphite - the first years. J Nucl Mater 1981, 100:55-63.
44. Kim Y.S., Hofman G.L., Robinson A.B., Snelgrove J.L., Hanan N. Oxidation of aluminum alloy cladding for research and test reactor fuel. J Nucl Mater 2008, 378(2):220-228.
45. Leenaers A., Detavernier C., Van den Berghe S. The effect of silicon on the interaction between metallic uranium and aluminum: a 50 year long diffusion experiment. J Nucl Mater 2008, 381:242-248.
46. Greenfield P. A comparison of Magnox A 12 and ZA alloys as canning materials for nuclear reactors. J Nucl Mater 1964, 11:121-134.
47. Popple L.A. The oxidation of magnesium alloys in reactor atmospheres. J Nucl Mater 1963, 8:60-76.
48. Reactor design. [accessed 19.07.10]. http://www-diva.eng.cam.ac.uk/energy/parksteaching/nuclear/4a1reactordesign.pdf.
49. Lobb R.C., Jones R.B. The influence of iodine vapour on creep rupture properties of 20%Cr/25%Ni/Nb stabilised stainless steel. J Nucl Mater 1976, 59:280-292.
50. Lobb R.C. The effect of iodine vapour on creep rupture properties of nitrided 20Cr-25Ni-Nb-1.5 Ti stainless steel. J Nucl Mater 1978, 74:212-220.
51. Antill J.E., Peakall K.A., Smart E.F. Corrosion of stainless steel in the presence of caesium. J Nucl Mater 1975, 56:47-60.
52. Gravenor J.G., Douglas J. Elevated temperature tensile properties of irradiated 20/25/Nb stainless steel fuel pin cladding at low and high strain rates. J Nucl Mater 1988, 158:87-95.
53. Yamada H. Failure of type 316 stainless steel cladding during thermal transients. J Nucl Mater 1978, 78:24-32.
54. Walker S.P., Yu A., Fenner R.T. Pellet-clad mechanical interaction: pellet-clad bond failure and strain relief. Nucl Eng Des 1992, 138:403-408.
55. Olander DR, Motta AT. Light water reactor materials. Unpublished work; 2009.
56. Cox B., Wong Y.M. A hydrogen uptake micro-mechanism for Zr alloys. J Nucl Mater 1999, 270:134-146.
57. Le Saux M., Besson J., Carassou S., Poussard C., Averty X. Behavior and failure of uniformly hydrided Zircaloy-4 fuel claddings between 25°C and 480°C under various stress states, including RIA loading conditions. J Eng Fail Anal 2010, 17(30):683-700.
58. Cox B. Some thoughts on the mechanisms of in-reactor corrosion of zirconium alloys. J Nucl Mater 2005, 336:331-368.
59. Motta A.T., Silva M.J.G., Yilmazbayhan A., Comstock R.J., Cai Z., Lai B. Microstructural characterization of oxides formed on model Zr alloys using synchrotron radiation. J ASTM Int 2008, 5(3).
60. Cox B. Pellet-clad interaction (PCI) failures of zirconium alloy fuel cladding - a review. J Nucl Mater 1990, 172:249-292.
61. Bossis P., Pecheur D., Hanifi K., Thomazet J., Blat M. Comparison of high burn-up corrosion on M5 and low tin zircaloy-4. J ASTM Int 2006, 3(1).
62. Sabol G.P. ZIRLO™ - an alloy development success. J ASTM Int 2005, 2(2).
63. Duplin N., Ansara I., Servant C., Toffolon C., Lemaigna C., Brachet J.C. A thermodynamic database for zirconium alloys. J Nucl Mater 1999, 275:287-295.
64. R. J. Pérez, Thermodynamic database for zirconium alloys, PhD thesis. KTH, Stockholm; 2006.
65. Was G.S., Ampornrat P., Gupta G., Teysseyre S., West E.A., Allen T.R., et al. Corrosion and stress corrosion cracking in supercritical water. J Nucl Mater 2007, 371:176-201.
66. Zinkle S.J., Busby J.T. Structural materials for fission & fusion energy. Mater Today 2009, 12(11):12-19.
67. Degueldre C., Gueneau Ch. Introducing the nuclear material challenges. J Nucl Mater 2006, 352:ix-xiii.
68. Murty K.L., Charit I. Structural materials for Gen-IV nuclear reactors: challenges and opportunities. J Nucl Mater 2008, 383(1-2):189-195.
69. Olander D.R. The Materials of fast breeder reactors. Naturwissenschaften 1980, 67:61-71.
70. Generation IV technology roadmap. [accessed 19.07.10]. http://gif.inel.gov/roadmap/pdfs/gen_iv_roadmap.pdf.
71. Yvon P., Carré F. Structural materials challenges for advanced reactor systems. J Nucl Mater 2009, 385:217-222.
72. Abram T., Ion S. Generation-IV nuclear power: a review of the state of the science. Energy Policy 2008, 36(12):4323-4330.
73. Fazio C., Alamo A., Almazouzi A., De Grandis S., Gomez-Briceno D., Henry J., et al. European cross-cutting research on structural materials for Generation IV and transmutation systems. J Nucl Mater 2009, 392:316-323.
74. Möslang A., Wiss T. From fission towards fusion. Nat Mater 2006, 5(9):679-680.
75. Workshop on advanced computational materials science: application to fusion and generation IV fission reactors. [accessed 19.07.10]. http://www.er.doe.gov/bes/reports/files/ACMS_rpt.pdf.
76. Baindur S. Materials challenges for the supercritical water-cooled reactor (SCWR). Bull Can Nucl Soc 2008, 29(1):32-38.
77. Ehrlich K., Konys J., Heikinheimo L. Materials for high performance light water reactors. J Nucl Mater 2004, 327:140-147.
78. Reiss T., Csom Gy., Fehér S., Czifrus Sz. The simplified supercritical water-cooled reactor (SSCWR), a new SCWR design. Prog Nucl Energy 2010, 52:177-189.
79. Johnston W.G., Rosolowski J.H., Turkalo A.M. Nickel-ion bombardment of annealed and cold-worked type 316 stainless steel. J Nucl Mater 1973, 48:330-338.
80. Kurishita H., Muroga T., Watanabe H., Yoshida N., Kayano H., Hamilton M.L. Effect of FFTF irradiation on tensile properties of P- and Ti-modified model austenitic alloys with small amounts of boron. J Nucl Mater 1994, 212-215:519-524.
81. Schafer L. An austenitic steel remaining ductile after neutron irradiation. J Nucl Mater 1996, 231:63-68.
82. Sagaradze V.V., Lapin S.S., Kirk M.A., Goshchitskii B.N. Influence of high-dose Kr+ irradiation on structural evolution and swelling of 16Cr-15Ni-3Mo-1Ti aging steel. J Nucl Mater 1999, 274:287-298.
83. Klueh R.L., Nelson A.T. Ferritic/martensitic steels for next-generation reactors. J Nucl Mater 2007, 371:37-52.
84. Lu Z., Faulkner R.G., Morgan T.S. Formation of austenite in high Cr ferritic/martensitic steels by high fluence neutron irradiation. J Nucl Mater 2008, 382:223-228.
85. Lu Z., Faulkner R.G., Flewitt P.E.J. Irradiation-induced impurity segregation and ductile-to-brittle transition temperature shift in high chromium ferritic/martensitic steels. J Nucl Mater 2007, 367-370:621-626.
86. Kurtz R.J., Alamo A., Lucon E., Huang Q., Jitsukawa S., Kimura A., et al. Recent progress toward development of reduced activation ferritic/martensitic steels for fusion structural applications. J Nucl Mater 2009, 386-388:411-417.
87. Yano Y., Yamashita S., Ohtsuka S., Kaito T., Akasaka N., Shibayama T., et al. Mechanical properties and microstructural stability of 11Cr-ferritic/martensitic steel cladding under irradiation. J Nucl Mater 2010, 398:59-63.
88. Rowcliffe A.F., Mansur L.K., Hoelzer D.T., Nanstad R.K. Perspectives on radiation effects in nickel-base alloys for applications in advanced reactors. J Nucl Mater 2009, 392:341-352.
89. Boothby R.M. The microstructure of fast neutron irradiated Nimonic PE 16. J Nucl Mater 1996, 230:148-157.
90. Angeliu T.M., Ward J.T., Witter J.K. Assessing the effects of radiation damage on Ni-base alloys for the Prometheus space reactor system. J Nucl Mater 2007, 366:223-237.
91. 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 Nucl Mater 2005, 340:93-112.
92. Mason R.E., El-Genk M. Experimental investigation of the ruthenium-uranium and rhenium-uranium binary systems. J Nucl Mater 1994, 217:304-321.
93. Ukai S., Fujiwara M. Perspective of ODS alloys application in nuclear environments. J Nucl Mater 2002, 307-311:749-757.
94. Ukai S., Hatakeyama K., Mizuta S., Fujiwara M., Okuda T. Consolidation process study of 9Cr-ODS martensitic steels. J Nucl Mater 2002, 307-311:758-762.
95. Wright I, Pint B, Lu ZP. Overview of ODS alloy development. [accessed 19.07.10]. http://www.ms.ornl.gov/FEM19/Proceedings/papers/Session%20II/wright.pdf.
96. Goshchitskii B.N., Sagaradze V.V., Shalaev V.I., Arbuzov V.L., Tian Y., Qun W., et al. Structure, radiation resistance and thermal creep of ODS ferritic steels. J Nucl Mater 2002, 307-311:783-787.
97. Certain AG, Field KG, Allen TR, Miller MK, Bentley J, Busby JT. Response of nanoclusters in a 9Cr ODS steel to 1 dpa, 525 °C proton irradiation, J Nucl Mater, in press. doi:10.1016/j.jnucmat.2010.07.002 [accessed 19.07.10].
98. Youshida E., Kato S. Sodium compatibility of ODS steel at elevated temperature. J Nucl Mater 2004, 329-333:1393-1397.
99. Schroer C., Konys J., Furukawa T., Aoto K. Oxidation behaviour of P122 and a 9Cr-2W ODS steel at 550°C in oxygen-containing flowing lead-bismuth eutectic. J Nucl Mater 2010, 398:109-115.
100. Takaya S., Furukawa T., Inoue M., Fujisawa T., Okuda T., Abe F., et al. J Nucl Mater 2010, 398:132-138.
101. Allen TR, Crawford DC. Lead-cooled fast reactor systems and the fuels and materials challenges. Science and Technology of Nuclear Installations, vol. 2007, Article ID 97486, pp. 11. doi:10.1155/2007/97486.
102. David Carpenter. Comparison of pellet-cladding mechanical interaction for zircaloy and silicon carbide clad fuel rods in pressurized water reactors. [accessed 19.07.10]. http://ocw.mit.edu/courses/nuclear-engineering/22-314j-structural-mechanics-in-nuclear-power-technology-fall-2006/projects/carpenter.pdf.
103. Katoh Y., Snead L.L., Henager C.H., Hasegawa A., Kohyama A., Riccardi B., et al. Current status and critical issues for development of SiC composites for fusion applications. J Nucl Mater 2007, 367-370:659-671.
104. Meyer MK, Fielding R, Gan J. Fuel development for gas-cooled fast reactors. [accessed 19.07.10]. http://www.inl.gov/technicalpublications/Documents/3394992.pdf.
105. Ueta S., Aihara J., Yasuda A., Ishibashi H., Takayama T., Sawa K. Fabrication of uniform ZrC coating layer for the coated fuel particle of the very high temperature reactor. J Nucl Mater 2008, 376:146-151.
106. Yang Y., Dickerson C.A., Swoboda H., Miller B., Allen T.R. Microstructure and mechanical properties of proton irradiated zirconium carbide. J Nucl Mater 2008, 378:341-348.
107. Wongsawaeng D. Performance modeling of deep burn TRISO fuel using ZrC as a load-bearing layer and an oxygen getter. J Nucl Mater 2010, 396:149-158.