No abstract

There are now twenty commercial nuclear power reactors operating as of May 2010 in South Korea. As nuclear capacity becomes higher and installations age, the Korean government and industry have launched R&D to estimate appropriate decommissioning costs of power reactors. In this paper, MCNP/ORIGEN2 code system which is being developed as a source term evaluation tool was verified by comparing the estimated nuclide inventory from MCNP/ORIGEN2 simulation with the measured nuclide inventory from chemical assay in an irradiated pressure tube discharged from Wolsong Unit 1 in 1994. Equilibrium core model of Wolsoung unit 1 was used as a neutron source to activate in-core and ex-core structural components. As a result, the estimated values from the analysis system agreed with measured data within 20% difference. Therefore, it can be concluded that MCNP/ORIGEN system could be a reliable tool to estimate source terms of decommissioning wastes from CANDU reactor, although this system assumes constant flux irradiation and snapshot equilibrium core model as a reference core.

To simultaneously vitrify Ion Exchange Resin, Zeolite, and Dry Active Waste generated from Korean Nuclear Power Plant, a borosilicate glass system was formulated. Viscosity and electrical conductivity of the glass were measured and within the desired ranges at the processing temperature. Those activation energies were evaluated as 152 and 70.46 kJ/mol within a temperature range of 1223 to 1623 K, respectively. Time-Temperature-Transformation study was performed using data from heat treatment. The hematite crystal was found within a temperature range of 823 to 1123 K. M äossbauer spectroscopy showed about 42 Fe 2＋ state existed in the glass produced from operation of the pilot-scale plant. Product Consistency Test performed from 7 to 120 d in the glass showed the leach rates of B, Na, Li and Si were much less than those of the benchmark glass. International Organization for Standardization test was performed at 363 K for 1022 d and shown that Cumulative Fraction Leached values of Na, Li, and Si were saturated below the fraction of 0.4 except that the leaching of B increases continuously. About 50 mm thickness layer was observed to be as a protective layer against continuous corrosion. According to Vapor Hydration Test, the corrosion rate of the glass was 2 g/m 2 /d and met the specification (50 g/m 2 /d). Based on Soxhlet leaching accomplished at 371 K for 30 d, the weight loss of the glass was determined as 106.8 g/m 2 which was lower than those of other HLW (High-Level radioactive Waste) glasses.

Glasses developed for the treatment of low-and intermediate-level radioactive waste (LILW) from nuclear power plants were evaluated by using the Material Characterization Center-1 (MCC-1) leaching method. Tests were conducted at temperatures of 40, 70, and 90 C for three weeks in pH buffer solutions spanning the range from pH 4 to pH 11. Normalized mass losses and forward dissolution rates of major glass elements (B, Na, Al, Si, Co, Cs) were analyzed under each leaching condition. From these data, the forward rate equations depending on pH and temperature were defined using a nonlinear regression method. This equation provided an overall diagram of the leach rate with these parameters (i.e., pH and temperature). The forward dissolution rates of the glasses were found to have a V-shaped pH dependence. The glasses in the pH ranges were found to have a forward dissolution rate below 10 g/m 2 Ád, when the temperatures were between 40 and 90 C and the leachant condition was pH 4-11. Except for the DG2 glass, the minimum forward dissolution rate (0.01-1 g/m 2 Ád) was obtained at approximately pH 7-8. Compared with previously reported results, the developed glasses showed relatively high forward dissolution rates at the neutral region, while showing similar or lower rates compared with other glasses and ceramic waste forms at both extremes of pH.

The Ulchin Vitrification Facility (UVF), to be used for the vitirification of low- and intermediate-level radioactive waste (LILW) generated by nuclear power plants (NPPs), is the world’s first commercial facility using Cold Crucible Induction Melter (CCIM) technology. The construction of the facility was begun in 2005 and was completed in 2007. From December 2007 to September 2009, all key performance tests, such as the system functional test, the cold test, the hot test, and the real waste test, were successfully carried out. The UVF commenced commercial operation in October 2009 for the vitrification of radioactive waste.