DEVELOPMENT OF A SYNTACTIC VALIDATION CAPABILITY FOR THE USE OF MCNP

MCNP has been widely used worldwide and has become an indispensable code to help solve challenging problems in many areas such as nuclear reactor analysis, medical physics, and criticality safety evaluation. MCNP provides a unique and rich syntax that allows users to model radiation transport problems with its capable geometry descriptors, flexible source definition, arbitrary material and nuclear data selection, and special tally (computation) options. MCNP's modeling flexibility comes from a myriad of cards, keywords, and operators which can be combined to describe geometry, set up physical conditions, or define quantities of interest (tallies). However, such a huge amount of syntax definitions presents challenges to MCNP beginners, or even intermediate-level users, to use them skillfully and correctly for problem definition and solution. Therefore, we have developed a syntactic validation capability for the use of MCNP. Such a capability can assist users to identify any syntax errors on-the-fly while preparing MCNP input files. It can also provide content assist, reference finding, and syntax highlighting functions for an accelerated user experience where error occurrence is proactively mitigated prior to input execution. This capability was implemented by developing a language server which can work with any Language Server Protocol-compatible text editor for MCNP input file development. In this paper, we describe the method development behind the language server and clarify how the syntactic validation works.