Neutronic Performance of (U, PU)C Fuel in a Lattice of Gfr Using Scale 6.0
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NEUTRONIC PERFORMANCE OF (U, Pu)C FUEL IN A LATTICE OF GFR USING SCALE 6.0 A. A. P. Macedo1, Carlos E. Velasquez1, C. A. M. da Silva1,2, C. Pereira1,2 1 Departamento de Engenharia Nuclear – Escola de Engenharia Universidade Federal de Minas Gerais, Av. Antônio Carlos, 6627, Pampulha, CEP 31270-901 Belo Horizonte, Tel/Fax: 55-31-34096662, MG, Brasil 2 Instituto Nacional de Ciências e Tecnologia de Reatores Nucleares Inovadores/CNPq, Brazil [email protected] ABSTRACT This paper studies the performance of (U, Pu)C fuel in a hexagonal assembly of a GFR (Gas Fast Reactor). The SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation version 6.0) code was used in the calculation. The goal is to evaluate the behavior of the infinite multiplication factor (kinf) for a heterogeneous assembly model using four nuclear data libraries: V6-238, V7-238, ENDF/B-VI.8 and ENDF/B-VII.0. The burnup of (U, Pu)C was performed by the TRITON-6 module, and the isotopic concentrations were evaluated during the cycle. The present work comprises calculations at Zero Power and Full Power condition. This study intends to achieve more information about different Fast Reactors.
INTRODUCTION In recent decades, nuclear power generation industries have focused on the improvement and development of reactor technology. In 2000, the IV Generation International Forum (GIF), presented three fast reactor concepts: the GFR (Gas-cooled Fast Reactor), the SFR (Sodium Fast Reactor) and the LFR (Lead-cooled Fast Reactor). The emphasis on these types of reactors is mainly due to the efficient use of natural resources, the possibility of reprocessing, the suitable management of radioactive waste and the operation for longer periods than conventional thermal reactors without requiring replenishment [9,10]. The fuel studied is uranium-plutonium carbide– (U, Pu)C in a ratio of 64/16/20 %, respectively. The cladding consists of SiC (50/50)%; the coolant is helium and the assembly has a hexagonal geometry [3]. The goal of this work is to evaluate the behavior of the infinite multiplication factor (kinf) for heterogeneous assembly model using four libraries of the code: V6-238, V7-238, ENDF/B-VI.8 and ENDF/B-VII.0. In this work is used SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation – version 6.0) code to simulate the hexagonal fuel assembly from a typical GFR. The geometry and the isotopic composition of the components were based on previous studies [1, 3]. The burn-up was performed using the TRITON-6 module. The isotopic concentrations were evaluated during the cycle. It was performed the evaluation of the (U, Pu)C fuel at Zero Power (ZP) and Full Power (FP) conditions.
METHODOLOGY General description of SCALE 6.0 SCALE 6.0 is a modular nuclear code developed by the Oak Ridge National Laboratory (ORNL) that has been used more frequently in recent years. The functional modules of SCALE include basic physics codes applicable to criticality, shielding fuel depletion, and radiation transport. The control modules of SCALE operate as s
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