Burnup calculations for accelerator driven thorium reactor using serpent monte carlo code

28164

In this work, a number of simulations have been performed in order to obtain neutronic and burnup characteristics of Accelerator Driven Thorium Reactor (ADTR) using Serpent Monte Carlo Code. This code is optimized for reactor lattice calculations. Center of ADTR model has a cylindrical natural Pb-Bi target, has 90 hexagonal structure fuel bundle in core and each bundle consist of 91 fuel pin. The length and radius of the target in the center of the system are 60 cm and 15 cm, respectively. The distribution of fuel rods consist of two types of fuel material with 232^Th and 233^U. The burnup calculations have been carried out using matrix exponential solution based on the Chebyshev Rational Approximation Method (CRAM) and the evaluated nuclear data used in the calculations was based on ENDF/B-6.8, ENDF-7, JEF-2.2, JEFF-3.1. The average power density is 38.6E-3 kW/g in designed system. Changes between the different nuclear cross section libraries were examined and observed to be compatible. All burnup calculations made up of 41 different steps up to 40 MW / kgU.

309

Monte Carlo Method , Thorium , Burnup , Subcritical Reactor

Erciyes University Journal Of The Institute Of Science an‎d Technology

314

Erciyes University Journal Of The Institute Of Science an‎d Technology