MCNP CALCULATION OF THE NEUTRON-PHYSICAL CHARACTERISTICS OF A REACTOR AND RADIATION PROTECTION IN A NUCLEAR POWER FACILITY FOR USE IN SPACECRAFT

摘要

MCNP [1] is a high-precision computer code that makes it possible to perform Monte Carlo simulation of neutron, γ-ray, and electron transport in systems with a complicated three-dimensional geometry. The code was developed over a period of many years, during which the range of solvable problems expanded. MCNP employs an ENDF/B [2] based library of nuclear constants with continuous variation of the particle energy.