Generation of Consistent Nuclear Properties of DUPIC Fuel by DRAGON with ENDFIBVI Nuclear Data Library

Abstract

DRAGON code with 89-groups ENDFB-VI cross section library was used in this paper to generate consistent nuclear properties of DUPIC fuel. The reference feed material used for the DUPE fuel cycle is a 17x17 French standard 900 MWe PWR spent fuel assembly with 3.2 w/o initial enrichment and 32500 MWD/T discharge burnup. The PWR fuel assembly was modeled by JPMT/SYBILT transport method in DRAGON to generate nuclide fields of spent PWR fuel. The resultant nuclide fields constitute the initial fuel composition files for reference DUPIC fuel which can be accessed by DRAGON for CANDU 2D cluster geometry depletion calculation and 3D supercell calculation. Because of uneven spatial power distribution in PWR assemblies and full core, unexpected transition cycle, and various fuel management strategy, the spent PWR fael composition is expected to be different from one assembly to the next. This heterogeneity was characterized also by modeling various spent PWR fuel assembly types in the paper. I INTRODUCTION The concept of the DUPIC (Direct Use of Spent PWR fuel in CANDU reactor) fuel cycle is to reuse the spent PWR fuel in CANDU reactor by an oxidation and reduction of oxide fuel (OREOX) process, which is technically feasible and safeguardable.' The study of DUPIC fuel cycle in CANDU reactor requires the pre- calculation of few group homogenized cross sections of DUPIC cluster cell and in-core reactivity devices such as adjuster rods and Zone Control Units(ZCUs). Because the fresh DUPIC fuel is made of spent PWR fuel, not only the typical CANDU cluster cell and supercell but the PWR assembly have to be modeled by a lattice code. In the previous studies~ the nuclear properties of DUPIC fuel were generated by different lattice code with various libraries: the composition of spent PWR fuel was calculated by CASMO-3 based on ENDFB-N, the base cross sections of DUPIC cell was obtained by WIMS-AECL based on ENDFB-V, and the incremental cross sections of in-core reactivity devices was generated by separate 3D transport code SHETAN. In addition, the ORIGEN code had to be used to link the CASMO-3 and WIMS-AECL codes. The advanced lattice analysis code DRAGON^ was designed for general geometry and can analyze both CANDU clusters and PWR assemblies. It also has the capability to do three-dimensional supercell transport calculations. It contains three modules for self-shielding calculation and transport calculation: (1) JPMT module: interface current method applied to homogeneous blocks (2) SYBILT module: collision probability method for simple ID or 2D geometry and the interface current method for 2D Cartesian or hexagonal assemblies (3) EXCELT module: collision probability method for more general 2D geometry and for 3D assemblies. To overcome the drawbacks of using inconsistent computational codes and corresponding libraries of previous studies, the DRAGON code with an 89-groups ENDFB-VI cross section library was used in this paper to generate consistent nuclear properties of DUPIC fuel in three steps: