The development of an MCNP tally-based burnup code Online publication date: Mon, 13-Jul-2009
by Bilal El Bakkari, Tarek El Bardouni, Ossama Merroun, Chafik El Younoussi, Yassine Boulaich, El Mahjoub Chakir
International Journal of Nuclear Energy Science and Technology (IJNEST), Vol. 4, No. 3, 2009
Abstract: The aim of this study is to evaluate the potential capabilities of a newly developed burnup code called BUCAL1. The code provides the full capabilities of the Monte Carlo code MCNP5 through the use of MCNP tally information. BUCAL1 uses the fourth-order Rung Kutta method with the predictor-corrector approach as an integration method to determine fuel composition at a desired burnup step. The validation of BUCAL1 is performed by code versus code comparison. The results of two different kinds of codes are employed. The first one is CASMO-4, a deterministic multigroup two-dimensional transport code. The second kind is MCODE and MOCUP, a link MCNP-ORIGEN codes. These codes use different burnup algorithms to solve the depletion equations system. The eigenvalue and isotope concentrations were compared for two PWR uranium and thorium benchmark exercises at cold (300°K) and hot (900°K) conditions, respectively.
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