Sodium fast reactor evaluation: Core materials Jin Sik Cheon a, * , Chan Bock Lee a , Byoung Oon Lee a , J.P. Raison b , T. Mizuno c , F. Delage d , J. Carmack e a Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong, Daejeon 305-353, Republic of Korea b Institute of Transuranium Elements, Karlsruhe, Germany c Japan Atomic Energy Agency, Oarai, Japan d Commission of Energie Atomic, Cadarache, France e Idaho National Laboratory, Idaho Falls, ID 83415-3860, USA article info abstract In the framework of the Generation IV Sodium Fast Reactor (SFR) Program the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. In this paper the status of available and developmental materials for SFR core cladding and duct applications is reviewed. To satisfy the Generation IV SFR fuel requirements, an advanced cladding needs to be devel- oped. The candidate cladding materials are austenitic steels, ferritic/martensitic (F/M) steels, and oxide dispersion strengthened (ODS) steels. A large amount of irradiation testing is required, and the compat- ibility of cladding with TRU-loaded fuel at high temperatures and high burnup must be investigated. The more promising F/M steels (compared to HT9) might be able to meet the dose requirements of over 200 dpa for ducts in the GEN-IV SFR systems. Ó 2009 Elsevier B.V. All rights reserved. 1. Introduction In the sodium fast reactors (SFR) envisioned in the Generation IV program, fuel pins are expected to be irradiated at a higher tem- perature to a higher burnup than in previously operated reactors. Core materials of concern here are fuel components such as the cladding and duct. Integrity of the fuel pins strongly depends on whether the cladding can withstand the irradiation environment. It is preferred that the cladding have a low swelling and high duc- tility throughout its lifetime. Sufficient creep strength at a high temperature is required. The materials for the duct must retain suf- ficient strength and toughness to allow duct handling after irradi- ation. This paper describes the main factors controlling the lifetime of these materials. The viability of the currently available materials is investigated for their utilization as the core materials of Gener- ation IV SFR systems. 2. Fuel rod cladding 2.1. Austenitic steels Austenitic steels were selected as the first materials for the cladding as well as for the duct of first generation fast reactors. Type 304 or 316 was used. These steels were chosen based on their corrosion resistance and good thermal creep resistance. Also, they have been favored in the sense of their high-temperature mechan- ical strength, good fabrication technology, and abundant experi- ence. However, they exhibited excessive swelling at doses above 50 dpa. This swelling was decreased by adding stabilizing ele- ments, by adjusting chemical composition, and by introducing cold work [1,2]. The irradiation behaviors of the austenitic steels have been summarized [2]. Radiation-induced void swelling is a life limiting factor for these steels when used for the fuel rod cladding of fast breeder reactors. The swelling increases with the neutron fluence at a given temperature. There is an incubation period after which swelling begins, as shown in Fig. 1. After the incubation period, steady-state swelling rate is almost constant, at 1%/dpa for austen- itic steels [3]. The swelling depends on the irradiation temperature, neutron flux, and applied stress [2]. As shown in Fig. 2, significant improvements have been made to reduce swelling by increasing the incubation period through adding stabilizing elements, varying chemical composition, and applying cold work [4]. Stainless steels are alloyed with Ti, B and P and cold worked by 20% for the PNC 316, 15-15Ti, and D9 austenitic steels. The chemical compositions are shown in Table 1. As the swelling increases, however, its influ- ence on the ductility becomes significant. Fig. 3 shows that, in the swelling range above 6%, these steels become too brittle to be handled [3,5,6]. The status of the development is summarized in Ref. [5]. Abun- dant irradiation experience has been accumulated. Titanium-stabi- lized and cold worked steels have been used; D9 in the US, 15-15Ti in France, DIN 1.4970 in Germany. These steels have good creep strength such that MOX fuel with D9 cladding was irradiated in 0022-3115/$ - see front matter Ó 2009 Elsevier B.V. All rights reserved. doi:10.1016/j.jnucmat.2009.03.021 * Corresponding author. Tel.: +82 42 868 2648; fax: +82 42 868 8709. E-mail address: jscheon@kaeri.re.kr (J.S. Cheon). Journal of Nuclear Materials 392 (2009) 324–330 Contents lists available at ScienceDirect Journal of Nuclear Materials journal homepage: www.elsevier.com/locate/jnucmat