SCK-CEN spent fuel libraries

This dataset is composed of a set of reference spent fuel libraries. The dataset contains the material composition and radiation emission (i.e. neutrons and gamma-rays) of spent fuel. The data was obtained with computer simulations using the ORIGEN-ARP code, which is part of the SCALE 6.1 package. Both PWR and BWR fuel geometries are included in the dataset, and for each geometry both UO2 and MOX fuel materials are considered. The dataset contains the information for spent fuel with a broad range of initial enrichment (UO2 fuel) or initial fissile content (MOX fuel), discharge burnup, and cooling times. For each simulation the neutron spectra, divided into contributions from (α,n) reactions, spontaneous fission, and total neutron emission, as well as total gamma-ray spectra are included. The neutron emission from selected isotopes is also reported, divided in contributions from (α,n) reactions and spontaneous fissions.

本数据集由多组参考乏燃料库构成。数据集涵盖乏燃料的材料组成与辐射发射情况（含中子与γ射线辐射）。该数据通过使用SCALE 6.1软件包中的ORIGEN-ARP程序开展计算机模拟获取。数据集包含压水堆（Pressurized Water Reactor, PWR）与沸水堆（Boiling Water Reactor, BWR）两种燃料构型，且每种构型均覆盖二氧化铀（UO2）与混合氧化物（MOX）两类燃料材料。数据集涵盖了初始富集度（针对二氧化铀燃料）或初始裂变物质含量（针对混合氧化物燃料）、卸料燃耗以及冷却时间覆盖范围广泛的各类乏燃料相关信息。每一次模拟均包含中子能谱（细分为(α,n)反应贡献、自发裂变贡献与总中子发射量）以及总γ射线能谱。此外，数据集还报告了选定同位素的中子发射情况，同样划分为(α,n)反应与自发裂变的贡献来源。