目的 基于高温气冷堆乏燃料元件的源项特征,针对不同工况下的乏燃料元件干法贮存设施进行辐射安全分析,使其满足GB 11806—2019和GB 18871—2002中相关要求。方法 结合高温气冷堆球形乏燃料元件中各核素的放射性水平确定其光子能谱,采用蒙特卡罗程序针对转运容器和干法贮存设施内的通风孔洞和屏蔽塞缝隙建立三维精细化屏蔽模型,对乏燃料转运和贮存2种典型工况开展辐射安全分析。结果 转运容器表面最大剂量率为1.42 mSv/h,采用二级台阶结构的贮存井屏蔽塞缝隙处最大剂量率为3.99 μSv/h,工作人员年受照剂量为7.98 mSv,均满足相关标准中的剂量限值要求。结论 本研究验证了高温气冷堆乏燃料干法贮存方案的辐射安全设计可靠性,为同类设施的工程设计提供了关键数据支撑。
Abstract
The work aims to analyze the radiation safety of dry storage facilities of spent fuel elements under different working conditions based on the source term characteristics of spent fuel elements in the high-temperature gas-cooled reactor (HTGR), to make them meet relevant requirements of GB 11806-2019 and GB 18871-2002. The photon energy spectrum of spherical spent fuel elements in the HTGR was determined in combination with the radioactive levels of each nuclide. Monte Carlo codes were used to establish three-dimensional refined shielding models for ventilation ducts and shielding plug gaps within transport casks and dry storage facilities. Radiation safety analyses were conducted for two typical scenarios: spent fuel transport and storage. The results indicated that the maximum dose rate on the surface of the transport cask was 1.42 mSv/h, and the maximum dose rate at the shielding plug gap of the storage well utilizing a two-stage step structure was 3.99 μSv/h. Furthermore, the annual effective dose for workers was 7.98 mSv. All parameters complied with the dose limit requirements specified in relevant standards. This study verifies the reliability of the radiation safety design for the HTGR spent fuel dry storage scheme and provides critical data support for the engineering design of similar facilities.
关键词
乏燃料元件 /
干法贮存 /
辐射安全
Key words
spent fuel elements /
dry storage /
radiation safety
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