Monte Carlo calculation of radiation damage in first wall of an experimental hybrid reactor Hacı Mehmet S ß ahin * Gazi U ¨ niversitesi, Teknik Eg ˘ itim Faku ¨ ltesi, Teknikokullar, Ankara 06503, Turkey Received 11 August 2006; received in revised form 26 April 2007; accepted 30 April 2007 Available online 12 July 2007 Abstract This work was focused on the neutronic calculation of the nuclear parameters (neutron spectrum, displacement per atom (DPA), gas production, tritium breeding ratio (TBR), nuclear heating) for structural materials in the first wall (FW) and fuel clad (made of ferritic/ martensitic steels, vanadium alloy, silicon carbide, copper alloy, and stainless steel) of an experimental hybrid reactor using the most current Monte Carlo Neutron-Particle Transport code MCNP5 1.4. Neutronic calculations were performed using a (DT) fusion driver hybrid reactor under a neutron wall loud of 2.25 MW/m 2 by full reactor power for one year. Obtained results were compared with three different data libraries (ENDF/B-V, ENDF/B-VI and CLAW-IV). TBR values in the reactor blanket for all investigated materials became greater than the minimum requirement (TBR > 1.05). Nuclear parameters like DPA, He-production and nuclear heating were considered as radiation damage limits for structural materials, copper alloy (Cu0.5Cr0.3Zr) showed better performance than all inves- tigated materials. Ó 2007 Elsevier Ltd. All rights reserved. 1. Introduction Nuclear fusion has an enormous potential to provide a safe, clean and unlimited energy source particularly the abundant fusion fuel, contrary to fission fuel and conven- tional fuels. Besides, it has environmental advantages com- pared to other energy sources. But, the commercial pure fusion reactors have not been expected to run in the short period. Because the fusion reactor have many problems such as plasma instability, high-material damage, high- radiation, sensitive fluid jet design, high-driver power and other complex problems (S ß ahin et al., 1986; Pease, 1992; Akansu and U ¨ nalan, 2002; U ¨ beyli, 2003). In this respect, a fusion–fission reactor namely hybrid reactor would be an appropriate solution for some of these problems. Fur- thermore, the main aim of the hybrid reactor is improve the neutronic and economical performances of fusion reac- tors by means of higher energy production and/or signifi- cant fissile fuel breeding. Some advantages and main structure of the hybrid blanket has been given in earlier works (Conn et al., 1980; Greenspan, 1984; S ß ahin et al., 1984). In general, the fusion–fission (hybrid) is a combination of the fusion and fission processes. The fusion plasma is surrounded with a blanket made of the fertile materials ( 238 U or 232 Th) to convert them into fissile materials ( 239 Pu or 233 U) by transmutation through the capture of the high-yield fusion neutrons. The fertile materials may also undergo a substantial amount of nuclear fission, espe- cially, under the irradiation of the high-energetic 14.1 MeV (D, T) neutrons. In a hybrid reactor system, a fusion bree- der can produce up to 30 times more fissile fuel than a fis- sion breeder per unit of energy (Greenspan, 1984, S ß ahin et al., 1997). In addition, some of the bred fissile material burns in the hybrid blanket in situ (S ß ahin and Yapıcı, 1998; S ß ahin et al., 1997, 2002a,b, 2003). Another potential of a hybrid reactor is to burn up minor actinides, which are produced as nuclear waste material in LWRs, with the help of high-energetic fusion neutrons and product 233 U, 238 Pu, 0306-4549/$ - see front matter Ó 2007 Elsevier Ltd. All rights reserved. doi:10.1016/j.anucene.2007.04.011 * Tel./fax: +90 312 212 43 04. E-mail addresses: mesahin@gazi.edu.tr, hm.sahin@hotmail.com www.elsevier.com/locate/anucene Annals of Nuclear Energy 34 (2007) 861–870 annals of NUCLEAR ENERGY