THE USE OF CAD BASED RADIATION TRANSPORT IN SUPPORT OF SNS: DAG-MCNP6 Validation

摘要

The Spallation Neutron Source (SNS) facility at Oak Ridge is the world's highest strength pulsed neutron beam facility. The design and operation of the facility depends upon the accurate simulation of the proton induced reactions in the beam target and much deeper into the device. Historically MCNP/X was the main Monte Carlo (MC) simulation package used, hence relying on the traditional Combinatorial Solid Geometry (CSG) package within. There are efforts to begin design simulations of the Second Target Station (STS) integrating with multi domain physics modelling from the start. The DAGMC (Direct Accelerated Geometry Monte Carlo) software library has been integrated with the MCNP6 package to form DAG-MCNP6, a version of MCNP6 that can transport particles through CAD geometries. This paper will focus on the discussion of the development of DAG-MCNP6 and the recent benchmarking activities using the SNS.