As part of deterministic safety analysis, neutronics analysis is applied in all stages of nuclear fuel cycle: transportation, fuel utilization in the reactor core, and storage. The main capabilities and focus areas of Neutronics Analysis and Fuel Group (NFG) are:

• reactivity induced reactor transient and accident analysis by 3D kinetics PARCS-RELAP/TRACE coupled code system,
• analysis of reactor pressure vessel fluence by the DORT neutron transport code,
• criticality safety analysis, including the analysis of burnup credit approach for spent nuclear fuel storage and cooling pools by KENO-VI (SCALE 6) and STARBUCS (SCALE 6) codes,
• development of neutron-nucleus interaction cross-section libraries by the HELIOS, TRITON (SCALE 6) codes,
• analysis of nuclear fuel depletion by ORIGEN (1D), TRITON (2D, 3D), MCNP 6.1 (3D), ORIGEN-KENO-VI (3D), MONTEBURNS (3D) codes,
• analysis of nuclear fuel thermal-mechanical behavior by TRANSURANUS code,
• spent fuel decay heat analysis by ORIGEN (SCALE 6) code, and
• shielding analysis (e.g. spent fuel cask, NPP containment rooms, liquid waste drums) by MORSE, MONACO/MAVRIC (SCALE 6) codes.

This group was established in 2003, and since then it has been successfully involved in regulatory review and independent safety analysis of Armenian Nuclear Regulatory Authority’s decision-making.