MCNPX–BUCAL1 code to code verification through burnup analysis B. El Bakkari a,b,⇑ , T. El Bardouni b , B. Nacir a , C. El Younoussi a,b , Y. Boulaich a,b , H. Boukhal b a Reactor Operating Unit (UCR), National Centre of Sciences, Energy and Nuclear Techniques (CNESTEN/CENM), POB 1382, Rabat, Morocco b ERSN–LMR, Department of Physics, Faculty of Sciences, POB 2121, Tetuan, Morocco article info Article history: Received 28 January 2013 Received in revised form 24 April 2013 Accepted 30 April 2013 Available online 6 June 2013 Keywords: MCNPX2.7 BUCAL1 Burnup Verification TRIGA reactor ENDF/B-VII.0 abstract The availability of accurate burnup data is an essential first step in any systematic approach to enhance- ment of economics, safety and performance of a research reactor. This first step requires the utilization of a well verified burnup code system. In this work a newly home-developed burnup code called BUCAL1 is presented. The code provides the full capabilities of the Monte Carlo neutron and photon transport code MCNP (version 5c). BUCAL1 has the capability of using several depletion calculation schemes that do not exist in several other burnup code systems such as: shuffling, refueling and multicycles burnup calcula- tion, in an automatic way. The accuracy and precision of BUCAL1 were tested for U-Zrh fuels, by a code to code verification with MCNPX2.7, by predicting the burnup parameters of the 2 MW TRIGA Mark II Moroccan research reactor. Continuous energy cross section data from the more recent nuclear data evaluation ENDF/B-VII.0 as well as S(a, b) thermal neutron scattering functions distributed with the MCNP code were used. Analysis of the verification results shows that BUCAL1 is enough accurate to be used in burnup calculations. Ó 2013 Elsevier Ltd. All rights reserved. 1. Introduction Several codes or combination of codes have been developed to perform Monte Carlo depletion analysis. In these burnup codes the main approach is to use neutron absorption and fission reac- tion information generated via neutronics codes to determine the nuclide composition at a next time step. This kind of model allows the integration of all neutron flux information into the calculation without post-processing and additional manipulation of neutron flux and cross-sections set. Neutron absorption and fission reaction data for individual nuc- lides are available as output from the Monte Carlo codes like MCNP through the use of tallies. The only requirement is point wise en- ergy-dependent cross section set which is available for each nu- clide of interest at required temperature. This paper presents a new home-developed burnup code, BU- CAL1, coupled with MCNP (version 5c) Monte Carlo code (X-5 Monte Carlo Team, 2003). MCNP can model extremely complex three-dimensional geometries. So, BUCAL1 is quite accurate over a given region because MCNP-generated reaction rates are inte- grated over the continuous-energy nuclear data and the space within the region. Thus, any oddly or regularly shaped region in MCNP can be depleted. The mean features of BUCLA1, regarding the existing other code systems, can be summed up on its capabil- ity to do several depletion schemes required for fuel management analysis, such as shuffling and refueling and the calculation of sev- eral important burnup parameters needed for reactor and safety analysis: criticality, thermal hydraulic, shielding, etc. The accuracy of BUCAL1 results were tested for different kinds of fuels (MOX, UO2 and ThO2-UO2) with different irradiation con- ditions in previous published papers (El Bakkari et al., 2009a,b; El Bakkari et al., 2012). In this work, MCNPX2.7 (Pelowitz, 2011) was chosen as a reference code to assess the accuracy of BUCAL1 for U-ZrH fuels, by predicting the burnup parameters of the 2 MW TRIGA Mark II Moroccan research reactor. Neutron cross sec- tion evaluations, based on the ENDF/B-VII (Chadwick et al., 2006) nuclear data library, were used. The ENDF/B-VI (McLane, 2001) was used for S(a, b) treatment. 2. BUCAL1 code system BUCAL1 is a FORTRAN computer code designed to aid in analy- sis, prediction, and optimization of fuel burnup performance in nu- clear reactors. The code was developed, at the Laboratory of Matter and Radiation (LMR) of University ABDELMALEK ESSADI Tetuan – Morocco, to incorporate the neutron absorption reaction tally infor- mation generated directly by MCNP(5c) code in the calculation of fissioned or neutron-transmuted isotopes for multi-fueled regions. This allows us to benefit of the full capabilities provided by MCNP and to incorporate them into burnup calculations in the aim to per- form more accurate and robust treatment of the problem. Neutron 0306-4549/$ - see front matter Ó 2013 Elsevier Ltd. All rights reserved. http://dx.doi.org/10.1016/j.anucene.2013.04.037 ⇑ Corresponding author at: Reactor Operating Unit (UCR), National Centre of Sciences, Energy and Nuclear Techniques (CNESTEN/CENM), POB 1382, Rabat, Morocco. Tel.: +212 668782702; fax: +212 537803326. E-mail address: bakkari@gmail.com (B. El Bakkari). Annals of Nuclear Energy 60 (2013) 242–247 Contents lists available at SciVerse ScienceDirect Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene