OpenMC Documentation
OpenMC project (MIT Computational Reactor Physics Group with Argonne National Laboratory)
Documentation for the open-source Monte Carlo neutron and photon transport code used across national labs and universities. Pairs a methods guide on transport theory with a Python API and notebooks, so you can run real criticality and depletion calculations.
More resources on Nuclear Engineering
An MCNP6 Primer
A Kansas State teaching primer, hosted by Los Alamos, for the MCNP radiation transport code. Worked input decks build up geometry, materials, sources, tallies and variance reduction, so you can set up shielding and dose calculations.
Perspectives on Reactor Safety (NUREG/CR-6042, Rev. 2)
The NRC Technical Training Center's own course text on nuclear safety concepts. Five modules trace defense in depth, severe accident progression in the vessel, containment behavior, and source terms, giving you the regulator's framework for accident analysis.
Nuclear Systems Volume I: Thermal Hydraulic Fundamentals, Third Edition
The reference text for nuclear thermal-hydraulics, used worldwide in graduate courses. Works through single- and two-phase flow, heat transfer, and boiling crisis so you can size coolant channels and check thermal margins in a core.
Nuclear Reactor Analysis
The standard graduate text on reactor physics, free in full from Michigan's institutional repository. Covers neutron transport, multigroup diffusion, criticality, reactor kinetics and core design, so you can set up and solve core-analysis problems yourself. Repeatedly named the best reactor-physics text; its multigroup diffusion and neutron-transport treatment is the benchmark.