A neutronic evaluation of the (Pu–U) and (Am–Pu–U) insertion in a typical fuel of Angra-I Antonella Lombardi Costa, Cláubia Pereira * , Maria Auxiliadora Fortini Veloso, Clarysson A.M. da Silva Departamento de Eng. Nuclear – Universidade Federal de Minas Gerais, PCA1, Anexo Engenharia, Av. Antônio Carlos, 6627, 31270-901 Belo Horizonte, MG, Brazil article info Article history: Received 23 June 2008 Received in revised form 7 November 2008 Accepted 10 November 2008 Available online 21 December 2008 abstract The goal is to evaluate the neutronic behavior when (Pu–U) and (Am–Pu–U) mixed oxide are inserted in a typical cell of a Pressurized Water Reactor (PWR) such as Angra-I. Four types of fuels were studied: (1) MOX fuel enriched at 3.1% and V m /V f = 1.15; (2) MOX fuel enriched at 4.5% and V m /V f = 1.15; (3) MOX fuel enriched at 4.5% and V m /V f = 2.0 and (4) MOX fuel enriched at 4.5%, with 1% of Americium insertion in its composition (62.8% of Am 241 , 0.1% of Am 242m and 37.1% of Am 243 ) and with V m /V f = 2.0. The first case rep- resents the standard state of Angra I, but with Pu. The second case is similar to the first but the enrich- ment is increased. To evaluate the Americium insertion, a study of the V m /V f was made and better results were obtained with V m /V f = 2.0 and to compare, this case was too evaluated to (Pu–U) in the third and fourth cases. The idea is to verify the possibility of using these fuels in Angra-I analyzing neutronic parameters such as infinite multiplication factor, hardening spectrum, Boron worth and reactivity tem- perature coefficients. The results show that it is possible to use all the studied fuels in Angra-I as well as to burn Am inserted in the MOX fuel by a considerable quantity during PWR operation. The WIMS- D5 code was used to perform a simplified neutronic and burnup simulations to evaluate this possibility. Ó 2008 Elsevier Ltd. All rights reserved. 1. Introduction The nuclear power industries have been considering, carefully, the future of the irradiated nuclear fuels discharged annually all over the world from the nuclear power plants. The nuclear waste materials are classified as fission products, structural materials and minor actinides (MA) defined as hazardous radioactive waste because of their long term, high level radiotoxicity. The MAs (Np 237 , Am 241 , Am 243 , etc.) are very important in nuclear waste management mainly because of the long term radiotoxicity due their extremely long half-lives. The strategies for the management of spent fuel in each country vary from reprocessing to direct disposal. In many countries, as in Brazil, most spent fuel is in storage at nuclear power plants, but if no decisions on management strategies are made, the large amount of spent fuel in storage will continue to increase (Interna- tional Atomic Energy Agency, 2007). The disposition of spent fuel can be either disposal in geological repositories or recycled, through reprocessing techniques. Therefore, the burned fuel reprocessing techniques, the trans- mutation of MAs and fission products and the geologic storage techniques are important tools to guarantee the future utilization of the nuclear energy as a more clean and safe source of energy generation (Proceedings of the Seventh Information Exchange Meeting, 2002). Minor actinides have been inserted in a point of the fuel cycle and they undergo transmutation or incineration dur- ing the reactor operation. Particularly, the Americium burnup in the thermal reactor cores has presented good results in its trans- mutation (Berthou et al., 2003; Costa and Pereira,2002,2003; Charl- ton et al., 1999; Kingdon et al., 1999; Takeda et al., 1998; Mori et al., 1997a,b; Kimura et al., 1994). Moreover, Americium transmutation in the fast reactors (Osaka et al., 2005), in the CANDU reactors (S ßahin et al., 2008), in Energy Amplifier Systems (Dahlfors et al., 2007) and in several other specific MA burners systems have been investigated. The main aim of this work is to evaluate the neutronic behav- iour resulting of the insertion of Pu–U and Am–Pu–U fuel in a PWR such as Angra-I. In the investigations, changes in the ratio V m /V f (moderator volume/fuel volume) as well as the fuel enrich- ment are considered. The simulations were performed using the WIMS-D5 nuclear code. The Winfrith Improved Multigroup Scheme (WIMS) is a general code for reactor lattice cell calculation on a wide range of reactor systems. Either the code models rod or plate fuel geometries in regular arrays or in clusters and the energy group structure has been chosen primarily for thermal calcula- tions. To verify the neutronic behavior of the core with Pu and Am insertion, the cell calculation is sufficient for the purposes of the present work. Future investigations can be performed using coupled codes such as, for example, WIMS-CITATION to study the total Angra-I core behavior. 0306-4549/$ - see front matter Ó 2008 Elsevier Ltd. All rights reserved. doi:10.1016/j.anucene.2008.11.006 * Corresponding author. Tel.: +55 31 34996686; fax: +55 31 34996660. E-mail address: claubia@nuclear.ufmg.br (C. Pereira). Annals of Nuclear Energy 36 (2009) 1–6 Contents lists available at ScienceDirect Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene