ALVAND 2.0

An advanced computational code for thermo-neutronic design and analysis of nuclear reactor cores

About ALVAND 2.0

ALVAND 2.0 is an advanced computational software for 3D steady-state thermo-neutronic analysis of reactor cores having single phase coolant flows. It consists of a variety of solvers for treating hexagonal and rectangular shaped reactors. Multigroup forward (direct) and backward (adjoint) diffusion equations are efficiently discretized via a number of nodal and finite element subroutines. The equations are then coupled to the single heated channel model for each assembly to calculate thermal-hydraulic quantities in all radial and axial nodes. The software has a variety of features including:

• optimized for hexagonal and rectangular cores with different symmetries (30^o, 45^o, 60^o, 90^o, 120^o, 180^o) and different boundary conditions;
• supporting direct feeding of x-sections, or indirect via PMAXS libraries;
• estimating reactor core’s significant parameters over the cycle, e.g. k-effective, group fluxes, power, boron concentration, kinetic parameters, reactivity coefficient of fuel and moderator temperature, Xe-135/Sm-149 inventory, temperature distributions within the core, peaking factors, fuel burn up etc.;
• ability to set piecewise working plan for the core (BOC-EOC) with different power levels and control rod positions;
• various options for solving linear/non-linear system of equations accelerated by a proper techniques as well as parallel processing. An “Automatic” mode is also embedded for the user convenience;
• capable of multi-cycle core simulation by shuffling fuel assemblies and/or exchanging fuel types;
• extensive and well-organized output files.

This code is applicable for reactor engineers, core designers, safety inspectors, and academics. Extensive V&V process is under action to ensure reliability of solutions. This include running the code for different semi-analytic and realistic problems, and comparing the outcome against benchmark values obtained via established codes e.g. DONJON, PARCS and other resources. The outputs prove good agreement with the reference data.

 Future developments mainly focus on enabling dynamic features as well as plug-ins to a number of external TH codes (like ARAS or PARS) to realize accident analysis and improved time-dependent core TH evaluations. More options e.g. pin power reconstruction, detector response estimation, and residual heat inclusion are also planned.