Nuclear reactor full-core deterministic computational schemes in transport theory

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We are focusing on full-core reactor physics codes required for simulation of operation, fuel management and safety analyses of third- and fourth-generation reactors. Neutron diffusion theory is currently used for this purpose, but reference calculations require a more accurate representation of transport phenomena. Based on recent advances related to the MOC and to the evolution of computer clusters, we can now propose the implementation of a family of rigorous full-core solution technique relying on the discrete ordinates method (for regular geometries) and on the MOC (for unstructured geometries). The MOC we propose to develop is an original contribution based on: 1. linear sources. Scattering and fission sources are assumed to vary linearly along each track segment crossing the domain. A parabolic order diamond differencing scheme is used to obtain the flux moments in each volume. Most current implementations are based on flat sources approximation and step-characteristic differencing. 2. scattering anisotropy. With the MOC solution, we propose to implement a variable-order Legendre expansion model for scattering anisotropy. 3. synthetic acceleration. The MOC already implemented in DRAGON4 code is accelerated using a second-order algebraic collapsing (AC) method limited to flat sources and isotropic scattering. We propose to replace the second-order AC method with a first-order AC method and to generalize its scattering representation to higher orders. 4. parallelism. We propose to use a capability of Fortran-2008 ISO standard named "coarray" to implement a massively-parallel solution technique adapted to the MOC and to the AC method. The DRAGON4 implementation of the MOC is currently not parallelized. Using the Unified Portable C (UPC) is an alternative massively-parallel solution technique for algorithms written in C language.