Author's Personal Copy Iron Phosphate Glass for Immobilization of 99 Tc

In most cases authors are permitted to post their version of the article (e.g. in Word or Tex form) to their personal website or institutional repository. Authors requiring further information regarding Elsevier's archiving and manuscript policies are encouraged to visit: 99 Tc (surrogated by Re) was immobilized in an iron phosphate glass. $1.1 mass% Re was retained, possibly dissolved, in the iron phosphate glass. The 7-day PCT normalized release of Re was <10 À2 g/m 2. Re concentration in the glass rapidly decreased with increasing melting temperature and duration. a b s t r a c t Technetium-99 (99 Tc) can bring serious environmental threats because of its long half-life (s 1/2 = $2.1 Â 10 5 years), high fission yield ($6%), and high solubility and mobility in the ground water. The high volatility makes it difficult to immobilize 99 Tc in continuous melters vitrifying 99 Tc-containing nuclear wastes in borosilicate glasses. This work explores a possibility of incorporating a high concentration of 99 Tc, surrogated by the non-radioactive Re, in an iron phosphate glass by melting mixtures of iron phosphate glass frits with 1.5–6 mass% KReO 4 at $1000 °C. The retention of Re achieved was $1.1 mass%. The normalized Re release by the 7-day Product Consistency Test was <10 À2 g/m 2. Surprisingly, the Re escaped from the melt within a short time of heating, especially when the temperature was increased. Therefore, 99 Tc volatilization would still be a challenging task for its immobilization in iron phosphate glasses. Technetium-99 (99 Tc), a fission product generated during the irradiation of 235 U-enriched fuel for commercial power generation or 239 Pu production for nuclear weapons, is present in the worldwide nuclear wastes. Because of its long half-life (s 1/2 = $2.1 Â 10 5 years) and high fission yield ($6%), 99 Tc bears serious environmental threats [1]. Moreover, the stable form of Tc in oxidizing environments , the pertechnetate anion, Tc(VII)O 4 À , is highly soluble and mobile in ground water, and thus easily migrates through geologic systems, resulting in potential harm to biospheres [2–5]. Vitrification technology is the preferred process for nuclear waste immobilization in the form of borosilicate glasses [6,7]. The primary concern in vitrifying nuclear wastes containing Tc is its low retention in the glass, which is caused by its high volatility at elevated temperatures [8,9]. The Tc solubility in a borosilicate glass is estimated to …