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Attila is a state-of-the-art radiation transport based code that is efficient, accurate and straightforward to use. Geometric information is input to the code as CAD drawings. Material property information is input as a cross section library. The user has full control over standard transport options and output reports.
In phase I of this work, a benchmark reactor problem is analyzed using the Attila and MCNP codes. Two libraries are used with Attila – one generated with WIMS-ANL and one generated with SCALE. Neutron fluxes calculated by MCNP and Attila at various energies and locations are in good agreement. Calculated values of keff do not agree as well, with Attila/WIMS-ANL based values diverging from MCNP values by up to 2.6 percent.
In phase II, Attila is used with a 22 group library to analyze depletion of a unit cell. The unit cell is sufficiently detailed to allow simulation of the various core operating modes. Results of depletion studies show that flux and number densities throughout the unit cell agree closely after an exposure history similar to the OSTR 30 year operational history, regardless of the size of time steps utilized in the depletion calculation.
Phase III work involves full core eigenvalue calculations on a radially zoned TRIGA core designed to mimic the current state of the OSU core. Global and local parameters are compared with measured values. Results of all three phases indicate that the Attila radiation transport code is well suited for TRIGA reactor simulation. |
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