Since safety analysis of high-temperature gas-cooled reactors (HTRs) has recently become an important focus, the development of computer codes for these types of analyses has gained equal significance. However, in comparison to light water reactors, there is considerably less experimental data available for HTR code validation. The work reported here takes advantage of additional reaction rate data recently made available from the HTR-PROTEUS experiments to perform validation of the Advanced Gas Reactor Evaluator (AGREE) computer code, which is currently being used for design and safety analysis of several advanced HTRs as part of the U.S. Department of Energy Advanced Reactor Development Program. Multigroup cross sections were generated for AGREE using the Monte Carlo code Serpent with the ENDF/B-VII.1 data library. Full-core Monte Carlo calculations were also performed with both the Serpent code and the MCNP code to provide a code-to-code comparison with the deterministic AGREE full-core calculation using Serpent cross sections. The eigenvalue, control rod worth, and neutron flux and power distributions are generally in good agreement between AGREE with both the experimental data and the full-core Monte Carlo calculations. However, as expected of a neutron diffusion code, some discrepancies in AGREE are observed, particularly in the fast flux spectrum in regions outside of the core primarily due to neutron streaming effects.