Soon Dal Park

The measurement of 129I for the cement and the paraffin solidified low and intermediate level wastes (LILWs), spent resin or evaporated bottom from the pressurized water reactor (PWR) nuclear power plants.

In this paper a relatively simple and low cost analysis procedure to apply to a routine analysis of (129)I in low and intermediate level radioactive wastes (LILWs), cement and paraffin solidified evaporated bottom and spent resin, which are produced from nuclear power plants (NPPs), pressurized water reactors (PWR), is presented. The (129)I is separated from other nuclides in LILWs using an anion exchange adsorption and solvent extraction by controlling the oxidation and reduction state and is then precipitated as silver iodide for counting the beta activity with a low background gas proportional counter (GPC). The counting efficiency of GPC was varied from 4% to 8% and it was reversely proportional to the weight of AgI by a self absorption of the beta activity. Compared to a higher pH, the chemical recovery of iodide as AgI was lowered at pH 4. It was found that the chemical recovery of iodide for the cement powder showed a lower trend by increasing the cement powder weight, but it was not affected for the paraffin sample. In this experiment, the overall chemical recovery yield of the cement and paraffin solidified LILW samples and the average weight of them were 67+/-3% and 5.43+/-0.53 g, 70+/-7% and 10.40+/-1.60 g, respectively. And the minimum detectable activity (MDA) of (129)I for the cement and paraffin solidified LILW samples was calculated as 0.070 and 0.036 Bq/g, respectively. Among the analyzed cement solidified LILW samples, (129)I activity concentration of four samples was slightly higher than the MDA and their ranges were 0.076-0.114 Bq/g. Also of the analyzed paraffin solidified LILW samples, five samples contained a little higher (129)I activity concentration than the MDA and their ranges were 0.036-0.107 Bq/g.

Separation of chlorine in a uranium compound by pyrohydrolysis and steam distillation, and its determination by ion chromatography

For the determination of chlorine in uranium compound, analytical methods by using a steam distillation and a pyrohydrolysis have been developed. The steam distillation apparatus was composed of steam generator, distilling flask and condenser etc. The samples were prepared with an aliquot of LiCl standard solution and a simulated spent nuclear fuel. A sample aliquot was mixed with a solution containing 0.2 M ferrous ammonium sulfate-0.5 M sulfamic acid 3 mL, phosphoric acid 6 mL and sulfuric acid 15 mL. The chloride was then distilled by steam at the temperature of until a volume of is collected. The pyrohydrolysis equipment was composed of air introduction system, water supply, quartz reaction tube, combustion tube furnace, combustion boat and absorption vessel. The chloride was separated from powdered sample which is added with accelerator, by pyrohydrolysis at the temperature of for 1 hour in a quartz tube with a stream of air of 1 mL/min supplied from the water reservoir at . The chlorides collected in each absorption solution by two methods was diluted to 100 mL and measured with ion chromatography to determine the recovery yield. For the ion chromatographic determination of chlorine in molten salt retained in a metal ingot, the chlorine was separated by means of pyrohydrolysis after air and dry oxidation, and grinding for the sample.