Aomi, M, Baba, T, Miyashita, T, Kamimura, K, Yasuda, T, Shinohara, Y, Takeda, T. 2008. Evaluation of hydride reorientation behavior and mechanical property for high burnup fuel cladding tube in interim dry storage. Journal of ASTM International 5(9):1–21.
ASTM B353-02. 2002. Standard specification for wrought zirconium and zirconium alloy seamless and welded tubes for nuclear service. West Conshohocken (PA): ASTM International. URL: https://www.astm.org
ASTM B811-13. 2013. Standard specification for wrought zirconium alloy seamless tubes for nuclear reactor fuel cladding. West Conshohocken (PA): ASTM International. URL: https://www.astm.org
Bel, JJP, Wickham, SM, Gens, RMF. 2006. Development of the Supercontainer design for deep geological disposal of high-level heat emitting radioactive waste in Belgium. Materials Research Society Symposia Proceedings 932:23–32.
Blokhin, DA, Chernov, VM, Blokhin, AI, Denim, NA, Sipachev, IV. 2012. Nuclear physical properties of zirconium alloys E110 and E635 under long-term neutron irradiation in VVER-1000 reactor. Inorganic Materials: Applied Research 3:124–128.
Deng, B, Campbell, TJ, Burris, DR. 1997. Hydrocarbon formation in metallic iron/water systems. Environmental Science & Technology 31:1185–1190.
Garzarolli, F, Stehle, H, Steinberg, E. 1996. Behavior and properties of zircaloys in power reactors: A short review of pertinent aspects in LWR fuel. ASTM STP 1295:12–32.
Guipponi, C. 2009. Effets de la radiolyse de l’air humide et de l’eau sur la corrosion de la couche d’oxyde du Zircaloy-4 oxydé [PhD thesis]. Lyon. In French.
Henrici-Olive, G, Olive, S. 1976. Fischer-Tropsch synthesis – molecular-weight distribution of primary products and reaction-mechanism. Angewandte Chemie International Edition 15:136–141.
Hillner, E, Franklin, DG, Smee, JD. 2000. Long‐term corrosion of Zircaloy before and after irradiation. Journal of Nuclear Materials 278:334–345.
IAEA. 1998a. Waterside corrosion of zirconium alloys in nuclear power plants. Vienna: International Atomic Energy Agency (IAEA). IAEA-TECDOC-996.
IAEA. 1998b. Durability of spent nuclear fuels and facility components in wet storage. Vienna: International Atomic Energy Agency (IAEA). IAEA-TECDOC-1012.
IAEA. 2006. Understanding and managing ageing of material in spent fuel storage facilities. Vienna: International Atomic Energy Agency (IAEA). IAEA Technical Reports Series 443.
Johnson, LH, McGinnes, DF. 2002. Partitioning of radionuclides in Swiss power reactor fuels. Switzerland: Nagra. Report No. 02-07.
Kaneko, S, Tanabe, H, Sasoh, M, Takahashi, R, Shibano, T, Tateyama, S. 2002. A study on the chemical forms and migration behavior of carbon-14 leached from the simulated hull waste in the underground condition. MRS proceedings 757:II3.8.
Kato, O, Tanabe, H, Sakuragi, T, Nishimura, T, Tateishi, T. 2014. Corrosion tests of Zircaloy hull waste to confirm applicability of corrosion model and to evaluate influence factors on corrosion rate under geological disposal conditions. Materials Research Society Symposia Proceedings 1665:195–202.
Kim, K. 2010. Estimation of carbon-14 in nuclear power plant gaseous effluents. EPRI Technical Report 1021106.
Kim, Y-J, Rebak, RB. 2009. Photo-electrochemistry of zirconium alloys in high temperature waste – A review. CORROSION 2009, paper no. 09417, 22–26 March 2009, Atlanta, GA, USA (NACE, Houston, TX).
Lefebvre, F, Lemaignan, C. 1997. Irradiation effects on corrosion of zirconium alloy claddings. Journal of Nuclear Materials 248:268–274.
Mogoda, AS. 1999. Electrochemical behaviour of zirconium and the anodic oxide film in aqueous solutions containing chloride ions. Thin Solid Films 357:202–207.
Motooka, T, Komatsu, A, Tsukada, T, Yamamoto, M. 2013. Pitting potential of Zircaloy-2 in artificial seawater under gamma-ray irradiation. ECS Transactions 53:25–32.
Motta, AT, Couet, A, Comstock, R. 2015. Corrosion of zirconium alloys used for nuclear fuel cladding. Annual Review Materials Research 45:18.1–18.33.
Smith, HD, Baldwin, DL. 1993. An investigation of thermal release of carbon-14 from PWR Zircaloy spent fuel cladding. Journal of Nuclear Materials 200:128–137.
Tait, WS. 1994. An introduction to electrochemical corrosion testing for practicing engineers and scientists. USA: Pair O Docs Publications.
Tupin, M, Hamann, J, Cuisinier, D, Bossis, P, Blat, M, Ambard, A, Miquet, A, Kaczorowski, D, Jamard, F. 2015. Understanding of corrosion mechanisms of zirconium alloys after irradiation: Effect of ion irradiation of the oxide layers on the corrosion rate. ASTM STP 1543:438–478.
Wang, P, Was, GS. 2015. Oxidation of Zircaloy-4 during in situ proton irradiation and corrosion in PWR primary water. Journal of Materials Research 30:1335–1348.
Yamaguchi, T, Tanuma, S, Yasutomi, I, Nakayama, T, Tanabe, H, Katsurai, K, Kawamura, W, Maeda, K, Kitao, H, Saigusa, M. 1999. A study on chemical forms and migration behavior of radionuclides in hull wastes. Proceedings of the 7th International Conference on Radioactive Waste Management and Environmental Remediation, ICEM; 1999 September 26–30; Nagoya, Japan. New York: American Society of Mechanical Engineers.