发明名称 Method of disposing of radioactive metal waste using melting decontamination
摘要 Disclosed is a method of disposing of radioactive metal waste using melting decontamination, including sorting radioactive metal waste generated in nuclear fuel processing or production facilities by predetermined sorting criteria, and charging sorted metal waste into a melting furnace so as to be melted; adding a impurity remover to the melt of the melting furnace to remove generated slag; pouring the melt having no slag into a mold to form an ingot; subjecting the ingot to gamma spectroscopy using a gamma spectrometer to measure gamma rays of U-235 (185.72 keV, 57.2%) among uranium isotopes, performing detector calibration using a certified reference material and self-absorption correction depending on the density of a medium using MCNP computer code, and calculating total radioactivity of the ingot from the quantified radioactivity and mass of U-235; and efficiently and rapidly determining whether the ingot subjected to radioactivity measurement satisfies a clearance limit.
申请公布号 US8796500(B2) 申请公布日期 2014.08.05
申请号 US201313870745 申请日期 2013.04.25
申请人 Kepco Nuclear Fuel Co., Ltd. 发明人 Cho Suk Ju;Lee Young Bae;Seol Jeung Gun;Kim Yong Jae
分类号 A62D1/00 主分类号 A62D1/00
代理机构 Rabin & Berdo, P.C. 代理人 Rabin & Berdo, P.C.
主权项 1. A method of disposing of radioactive metal waste using melting decontamination, comprising: 1) sorting radioactive metal waste generated in nuclear fuel processing or production facilities by predetermined sorting criteria and charging the sorted metal waste into a melting furnace so that the metal waste is melted; 2) adding a impurity remover to a melt of the melting furnace and removing generated slag; 3) pouring the melt having no slag into at least one mold thus producing an ingot; 4) subjecting the ingot to gamma spectroscopy using a gamma spectrometer to measure gamma rays of U-235 (185.72 key, 57.2%) among uranium isotopes, performing detector calibration using a certified reference material (CRM) and self-absorption correction depending on a density of a medium using a MCNP (Monte Carlo N-Particle) computer code, and calculating total radioactivity of the ingot from quantified radioactivity and mass of U-235; and 5) determining whether the ingot subjected to radioactivity measurement satisfies a clearance limit.
地址 Daejeon KR