Dosimetry and analysis of the delayed gamma rays emitted from the irradiated holder device used in Mo-۹۹ production facilities

The vital radioisotope ۹۹mTc, used in approximately ۸۰-۸۵% of diagnostic radiopharmaceuticals, is mostly obtained from the decay of Mo-۹۹. To provide ۱۰۰ Ci Mo-۹۹ (۶ days) per week for Iran's domestic nuclear medicine centers, an industrial plan of Mo-۹۹ production via fission of Low Enrichment Uranium targets in the Tehran Research Reactor has recently been evaluated. In this regard and with the aim of better conceptual knowledge for the site's solid waste management, two holders made of Al-Alloy ۶۰۶۱ and Stainless Steel-۳۱۶, which will be used to irradiate the targets, were investigated in this paper. The gamma spectrum emitted from the irradiated holder device and the corresponding dose rate were calculated by a proposed method that had been verified experimentally. When the Al-Alloy ۶۰۶۱ holder device was used, the calculation results indicated that the ۹۹Mo activity produced in the mini-plates was higher and the gamma dose rate of the activated holder device decreased faster.

کلیدواژه ها:

Dosimetry ، Gamma spectroscopy ، ۹۹Mo production ، Tehran Research Reactor ، solid waste management ، MCNPX