Radiation safety justification of spent nuclear fuel management technology for a reactor facility

Abstract

Research at the Institute of Atomic Energy of the National Nuclear Center of the Republic of Kazakhstan (IAE NNC RK) is focused on developing effective technologies for the management of spent nuclear fuel (SNF) from the BN-350 fast reactor. Among the approaches under consideration is a dilution and immobilization technology developed for irradiated uranium–graphite fuel from the Impulse Graphite Reactor (IGR). The process involves dry mixing of highly enriched uranium (HEU) fuel with depleted uranium dioxide, followed by immobilization through cementation. This study presents a comprehensive assessment of radiation conditions throughout the technological cycle of HEU fuel processing. Radiation field analysis at each processing stage is essential for establishing design requirements, optimizing facility operation, and ensuring personnel protection. Monte Carlo simulations performed using the MCNP code were employed to evaluate dose distributions and support radiation safety justification of the facility design. Based on the simulation results, several design modifications were implemented to improve shielding effectiveness and reduce occupational exposure. The proposed approach introduces a systematic methodology for radiation field assessment under conditions of incomplete information on fuel distribution, providing a conservative framework for safety analysis. Given the absence of direct international analogues for the management of irradiated HEU graphite fuel, the developed methodology offers both practical value for the IGR fuel immobilization project and potential applicability to future SNF management technologies for the BN-350 reactor.

Keywords: Spent Nuclear Fuel, Management, measured effective dose, radiation situation.