Abstract
The Advanced Random Ray Code (ARRC) is a high performance computing application capable of high-fidelity simulations of full core nuclear reactor models. ARRC leverages a recently developed stochastic method for neutron transport, known as The Random Ray Method (TRRM), which offers a variety of computational and numerical advantages as compared to existing methods. In particular, TRRM has been shown to be capable of efficient simulation of explicit three dimensional geometry representations without assumptions about axial homogeneity. To date, ARRC has utilized Constructive Solid Geometry (CSG) combined with a nested lattice geometry which works well for typical pressurized water reactors, but is not performant for the general case featuring arbitrary geometries. To facilitate simulation of arbitrarily complex geometries in ARRC efficiently, we propose performing transport directly on Computer-Aided Design (CAD) models of the geometry. In this study, we utilize the Direct-Accelerated Geometry Monte Carlo (DAGMC) toolkit which tracks particles on tessellated CAD geometries using a bounding volume hierarchy to accelerate the process, as a replacement for ARRC’s current lattice-based accelerations. Additionally, we present a method for automatically subdividing the large CAD regions in the DAGMC model into smaller mesh cells required by random ray to achieve high accuracy. We test the new DAGMC geometry implementation in ARRC on several test problems, including a 3D pincells, 3D assemblies, and an axial section of the Advanced Test Reactor. We show that DAGMC allows for simulation of complex geometries in ARRC that would otherwise not be possible using the traditional approach while maintaining solution accuracy.
Highlights
The Advanced Random Ray Code (ARRC) is a recently developed neutron transport code that is capable of simulating nuclear reactors in high-fidelity
ARRC’s traditional constructive solid geometry (CSG) input scheme relies on a lattice based hierarchy for efficient performance to be achieved. While this means that ARRC’s approach to geometry is not efficient for the general case, this method is at least a good choice for pressurized water reactor (PWR) geometries that feature reactors composed of a lattice of fuel assemblies, which themselves are composed of a lattice of fuel pins
As the Flat Source Regions (FSRs) decrease in size, the ARRC simulation will approach an eigenvalue that is identical to the multi-group reference solution that was generated with OpenMC
Summary
The Advanced Random Ray Code (ARRC) is a recently developed neutron transport code that is capable of simulating nuclear reactors in high-fidelity. ARRC is built upon the Random Ray Method (TRRM) of neutral particle transport, which is a stochastic method that shares many. Both TRRM and ARRC have been documented in a number of publications already [1,2,3,4], which include more information on the theory behind TRRM. In the present study we introduce new geometry modeling capabilities into ARRC with the following goals: 1. Application of computer-aided design (CAD) models in ARRC as a user-friendly and portable geometry representation as opposed to ARRC’s application-specific constructive solid geometry (CSG) model
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