DOI: 10.55176/2414-1038-2019-4-67-78
Authors & Affiliations
Bikeev A.S., Daichenkova Yu.S., Kalugin M.A., Oleynik D.S., Shkarovsky D.A.
National Research Centre "Kurchatov Institute", Moscow, Russia
Bikeev A.S. – Head of Software Laboratory, National Research Center "Kurchatov Institute".
Daichenkova Yu.S. – Engineer, National Research Center "Kurchatov Institute". Contacts: 1, Akademika Kurchatova pl., Moscow, Russia, 123182. Tel: +7 (499) 196-91-49; e-mail:
Kalugin M.A. – Scientific Director of the Kurchatov Nuclear Power Complex, Dr. Sci. (Tech.), National Research Center "Kurchatov Institute".
Oleynik D.S. – Head of Audit Laboratory, National Research Center "Kurchatov Institute".
Shkarovsky D.A. – Head of the Division of Benchmark Calculations of Nuclear Reactors, Cand. Sci. (Tech.), National Research Center "Kurchatov Institute".
Abstract
In recent years, new approaches have been developed to ensure radiation safety during carting spent nuclear fuel from nuclear power plants with VVER-1000 and VVER-1200 reactors, because the height of the fuel assembly and the initial enrichment of fuel have increased; operating modes of reactors have changed and average burnup in fuel assemblies has increased; new requirements about two safety barriers and prohibition on using liquid neutron protection has appeared. The new generation transport container is designed to cart 18 spent fuel assemblies of VVER-1000 and VVER 1200 reactors to long-term storage.
The main objective of the work is development and testing of the full-scale mathematical model of the new generation transport container for analysis of radiation safety using the Monte Carlo method, as well as performing numerical studies on the selection of optimal non-analog modeling parameters. Using the simple iteration method, the optimal values of the splitting surfaces for the non-analog modeling method “weight window” were determined.
The developed three-dimensional highly detailed model makes it possible to take into account the contribution to the equivalent dose rate from ionizing radiation sources in the top and bottom nozzles, reduce the number of conservative approximations, and, thus, increase the accuracy of radiation safety analysis. Increasing the accuracy of calculating the equivalent dose rate on the surface of the transport container may make it possible to cart spent fuel assemblies with a greater burnup.
Keywords
MCU-PD, Monte Carlo method, computer simulation, spent nuclear fuel, radiation safety, transport container
Article Text (PDF, in Russian)
UDC 621.039.51
Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants, 2019, issue 4, 4:8