Byung-Youn Min

Volatility and leachability of heavy metals and radionuclides in thermally treated HEPA filter media generated from nuclear facilities.

The purpose of the present study was to apply thermal treatments to reduce the volume of HEPA filter media and to investigate the volatility and leachability of heavy metals and radionuclides during thermal treatment. HEPA filter media were transformed to glassy bulk material by thermal treatment at 900°C for 2h. The most abundant heavy metal in the HEPA filter media was Zn, followed by Sr, Pb and Cr, and the main radionuclide was Cs-137. The volatility tests showed that the heavy metals and radionuclides in radioactive HEPA filter media were not volatilized during the thermal treatment. PCT tests indicated that the leachability of heavy metals and radionuclides was relatively low compared to those of other glasses. XRD results showed that Zn and Cs reacted with HEPA filter media and were transformed into crystalline willemite (ZnO·SiO(2)) and pollucite (Cs(2)OAl(2)O(3)4SiO(2)), which are not volatile or leachable. The proposed technique for the volume reduction and transformation of radioactive HEPA filter media into glassy bulk material is a simple and energy efficient procedure without additives that can be performed at relatively low temperature compared with conventional vitrification process.

VOLUME REDUCTION OF DISMANTLED CONCRETE WASTES GENERATED FROM KRR-2 AND UCP

As part of a fundamental study on the volume reduction of contaminated concrete wastes, the separation characteristics of the aggregates and the distribution of the radioactivity in the aggregates were investigated. Radioisotope 60 Co was artificially used as a model contaminant for non-radioactive crushed concrete waste. Volume reduction for radioactively contaminated dismantled concrete wastes was carried out using activated heavy weight concrete taken from the Korea Research Reactor 2 (KRR-2) and light weight concrete from the Uranium Conversion Plant (UCP). The results showed that most of the 60 Co nuclide was easily separated from the contaminated dismantled concrete waste and was concentrated mainly in the porous fine cement paste. The heating temperature was found to be one of the effective parameters in the removal of the radionuclide from concrete waste. The volume reduction rate achieved was above 80% for the KRR-2 concrete wastes and above 75% for the UCP concrete wastes by thermal and mechanical treatment.