Neutronics Modeling and Simulation of Sharp for Fast Reactor Analysis

Fast Reactor Physics and Computational Methods

“…A new multigroup cross section generation code MC 2 -3 [51,52] has recently been developed at ANL by improving the resonance self-shielding and spectrum calculation methods of MC 2 -2 and integrating the onedimensional cell calculation capabilities of SDX. MC 2 -3 generates multigroup cross sections for every cell type by solving a homogeneous medium or a heterogeneous slab or cylindrical unit cell problem in UFG (~2000) or HFG (~400,000) level.…”

“…[51,64] The application scope targeted for UNIC ranges from the homogenized assembly approach prevalent in current reactor analysis methodologies to explicit geometry, time-dependent transport calculations that are directly coupled to thermal-hydraulics and structural mechanics calculations in reactor accident simulations. Three transport equation solvers have been developed based upon the unstructured finite-element mesh representation: PN2ND, SN2ND, and MOCFE.…”

ARRC: A random ray neutron transport code for nuclear reactor simulation

“…As each crossing must be completed sequentially in order, the work to complete traversal of that cycle cannot be parallelized, resulting in an extremely long runtime for each power iteration. Additionally, as it is necessary to use domain decomposition when simulating large full core reactor problems on supercomputers [25,26,29,30,31] (as is discussed in subsection 2.5), traversal of full track cycles every power iteration may require that a ray is transmitted between subdomains handled by different processes many thousands of times each power iteration, which may incur impractical synchronization and network communication costs. 5.…”

Prospects in Deterministic Three-Dimensional Whole-Core Transport Calculations

“…A detailed discussion of the numerical schemes used in UNIC, as well as results for numerous benchmarks using the three flux solvers and analysis of parallel strategies and computation performance, can be found in Ref. (34). Most of the comparisons are usually done on the eigenvalue but in this reference a few comparisons are also given for reaction rates.…”

“…The equations of the method are positive definite and the one-group solution is efficiently obtained with a conjugate gradient technique. (33), (34) The SN2ND solver was developed for a better spatial detail with homogenization only at the pin level and was therefore based on the SN angular discretization which, for the same polynomial order, offers a higher degree of spatial accuracy than the PN2ND solver. Both solvers have been applied to the calculation of the homogenized ABTR reactors.…”