Jong-Kil Park

Waste minimization pretreatment via pyrolysis and oxidative pyroylsis of organic ion exchange resin

Abstract Pyrolysis and/or oxidative pyrolysis of organic ion exchange resins and other combustible waste may be effective pretreatment processes before vitrification. To further examine these processes, organic ion exchange resins were pyrolyzed or oxidatively pyrolyzed with the use of a small-scale, commercial TGA. Volatilization of the anionic and cationic resins was observed separately for each resin as a function of temperature for pyrolysis and oxidative pyrolysis conditions. The quantity of remains or residue was found to be dependent on the method employed. Three different methods were examined with the TGA to pretreat the resins: pyrolysis; oxidative pyrolysis; and oxidative pyrolysis of ash remaining after the pyrolysis of resin. The latter two methods were found to provide better volume reduction than the pyrolysis-only process. Between the two types of resins, cationic and anionic, the cationic exchange resin was less volatile. Pyrolysis and oxidative pyrolysis of mixed resin (50% cation and 50% anion by wt.) showed volatilization at the temperatures where volatilization was observed for each of the separate resins. Because of certain limitations of the commercial TGA, tube furnace experiments were performed, generally, to examine the pyrolysis of larger quantities of cationic, anionic, and mixed resin, and to examine off-gas characteristics. The cationic resin-only and anionic resin-only gravimetric results showed good agreement with the smaller-scale TGA results. SEM pictures of the different variants of the resin (cationic, anionic, and mixed) show a different morphology for each. Off-gas data showed the presence of H 2 S, SO 2 , CO, and NO during the pyrolysis of cationic resin. CO was observed during the pyrolysis of anionic resin. The mixed resin trials showed the presence of the gases approximately at the temperatures where the gases would evolve if the results of the two different resins (cationic and anionic) were superimposed. However, the amount of hydrogen sulfide relative to the sulfur dioxide was found to increase significantly compared to the results of the cationic resin-only trials.

냉각 방법에 따른 모의 방사성폐기물 유리고화체의 특성평가

In order to examine and compare the characteristics of two vitrified forms (AG8W1 and DG2) simulated for the operation of a commercial vitrification facility being constructed in Ulchin nuclear power plant, the vitrified forms were cooled by the natural cooling and annealing methods, respectively. And the Product Consistency Test (PCT), compressive strength, thermal conductivity, specific heat, phase stability, softening point and Coefficient of Thermal Expansion (CTE) of the vitrified forms were experimented. Consequently, it was shown that there were no significant differences on the physiochemical properties of the vitrified forms performed the natural cooling and annealing.

An experimental study on liquid radioactive waste treatment process using inorganic ion exchanger

Abstract Using commercial inorganic, organic sorbents and activated carbons, the treatment of low level liquid radwaste in nuclear power plant was studied. The batch, column and pilot tests were performed using simulants and actual liquid radwaste. The tests were focused on the removal capacity of the sorbents under sodium ion concentration since the liquid radwaste commonly contains a large amount of sodium ion which may affect the removal capacity of ion exchangers. The target ions to remove with sorbents were radionuclides such as Co‐58, Co‐60, Cs‐134, and Cs‐137 which are the major species in the liquid radwaste. The fundamental experiments showed that the inorganic sorbents and activated carbons have a better efficiency under sodium ion concentration for a cesium removal than a cobalt removal, whereas the organic resin is better for the cobalt removal. A new process for liquid radioactive waste treatment was proposed here based upon the experimental results. That is composed of a filter, activated ca...

Analysis of Leaching Behavior of Simulated LILW Glasses by Using the MCC-1 Test Method

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/m2·d, 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/m2·d) 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.

Characterization of Glass Melts Containing Simulated Low and Intermediate Level Radioactive Waste

In order to examine the process parameters for the vitrification of Low and Intermediate Level radioactive Waste (LILW) generated from nuclear power plants, measurements of several melt properties was performed for four selected glasses containing simulated waste. Electrical conductivity and viscosity were determined at temperatures ranging from 1123 to 1673℃. The temperature dependences of both properties in the molten state showed a similar behavior in which their values decrease as the temperature increases. The values of the electrical conductivity and viscosity at a temperature of 1423 K adopted in an induction cold crucible melter process were 0.27~0.42 S/cm and 9.8~42 dPas, respectively.

Glass Formulations for Vitrification of Low-and Intermediate-level Waste

In order to develop glass formulations for vitrifying Low-and Intermediate-Level radioactive Wastes (LILW) from nuclear power plants of Korea Hydro & Nuclear Power (KHNP) Co., Ltd., promising glass formulations were selected based on glass property model predictions for viscosity, electrical conductivity and leach resistance. Laboratory measurements were conducted to verify the model predictions. Based on the results, the models for electrical conductivity, US DOE 7-day Product Consistency Test (PCT) elemental release, and pH of PCT leachate are accurate for the LILW glass formulations. However, the model for viscosity was able to provide only qualitative results. A leachate conductivity test was conducted on several samples to estimate glass leach resistance. Test results from the leachate conductivity test were useful for comparison before PCT elemental release results were available. A glass formulation K11A meets all the KHNP glass property constraints, and use of this glass formulation on the pilot scale is recommended. Glass formulations K12A, K12B, and K12E meet nearly all of the processing constraints and may be suitable for additional testing. Based on the comparison between the measured and predicted glass properties, existing glass property models may be used to assist with the LILW glass formulation development.