Alpha activities can be used for categorization, transportation, and disposal of radioactive waste generated from the operation of nuclear facilities including nuclear power plants. In order to transport and dispose of such low- and intermediate-level radioactive waste (LILW) to the Wolsong LILW Disposal Center (WLDC) at Gyeongju, the gross alpha concentration of an individual drum should be determined according to the acceptance criteria. In addition, when the gross alpha concentration exceeds 10 Bq/g, the inventory of the comprising alpha emitters in the waste is to be identified. Gross alpha measurements using a proportional counter are usually straightforward, inexpensive, and high-throughput, so they are broadly used to assay the total alpha activity for environmental, health physics, and emergency-response assessments. However, several factors are thoughtfully considered to obtain a reliable approximate for the entire alpha emitters in a sample, which include the alpha particle energy of a particular radionuclide, the radionuclide that is used as a calibration standard, the uniformity of film in a planchet, time between sample collection and sample preparation, and time between sample preparation and counting. Korea Atomic Energy Research Institute (KAERI) have evaluated the inventory of radionuclides in low-level radioactive waste drums to send every year hundreds of them to the WLDC. In this presentation, we revisit the gross alpha measurement results of the drums transported to WLDC in the past few years and compare them with the concentrations of alpha emitters measured from alpha spectrometry and gamma spectrometry. This study offers an insight into the gross alpha measurement for radioactive waste regarding calibration source, self-absorption effect, composition of alpha emitters, etc.

Korea Atomic Energy Research Institute (KAERI) is planning to disposal of the radioactive contaminated cement waste form to the final disposal facility. The final disposal facility require evaluation of immersion, compressive strength, and radionuclide inventory of radioactive wastes to meet the acceptance criteria for safe disposal. According to the LILW acceptance criteria of the Nuclear Safety and Security Commission ok Korea (NSSC), the disposal limit radioactivity of 129I (3.70×101 Bq/g) is lower than other radionuclides. 129I emits low energy as its disposal limit is low, so it is difficult to analyze in the presence of 137Cs and 60Co which emit high energy. Therefore, it is essential to an accurately separate and analyze iodine in radioactive waste. In this study, we focused on the determination of 129I in cement waste form containing 137Cs, 60Co. We added 1 g of 129I(11.084 Bg), 137Cs(1,300 Bq) and 60Co(402 Bq) to cement waste form, respectively. The separation of 129I in cement waste form was carried out using an acid leaching method. And, we confirmed the specific activity of 137Cs and 60Co at each separation step. It was observed that an acid leaching method showed the remove efficiency 137Cs(99.97%) and 60Co(99.94%), respectively. In addition, 129I was also analyzed at approximately 96.44% in simulated contaminated cement waste form. In conclusion, through this experiment, it was confirmed that 129I could be successfully separated and analyzed by using the acid leaching method in cement waste form containing excessive 137Cs and 60Co.

We established pretreatment method of solidified cement ion-exchange resin samples generated before 2003 in nuclear power plants for measurement of non-volatile radionuclide activity. A microwave digestion system (MDS) with mixed acid (HCl-HNO3-HF-H2O2) was used to dissolve cement and to desorb non-volatile elements such as Ce, Co, Cs, Fe, Nb, Ni, Re, Sr and U from mixed ion-exchange resin. The content of Ce, Co, Fe, Nb, Ni, Re, Sr, U and Cs after pretreatment of cement plus mixed ion-exchange resin was measured by ICP-AES and ICP-MS, respectively. As iron and strontium are also present in cement, their content after dissolving a certain amount of cement was measured by ICP-AES. All elements except Nb were quantitatively recovered. Especially since the Nb recovery was low at 72.0±2.5%, the MDS following addition of the mixed acid to the resin was operated once more for desorbing Nb from it. Finally the recovery of Nb was over 95%. This sample pretreatment method will be applied to solidified cement ion-exchange resin samples generated in nuclear power plants for assessment of radionuclide inventory.

We conducted multi-elements determination of reference material certified by the Inorganic Ventures, IV-26, using iCAP 7400 ICP-OES of Thermo Fisher Scientific. And we statistically evaluated analysis results by introducing the in-house proficiency evaluation method implemented at the Ministry of Food and Drug Safety. Ca, Co, Fe, Mg, Ni, and V were selected as target elements, and extended uncertainty was estimated at a confidence level of about 95% and coverage factor k = 2. Five parameters incurred at manufacturing process (standard solution, calibration curve, repeated measurement and dilution factor of the test sample) were considered when determining the uncertainty. En-score can be calculated using the formula En=(x-X)/(Ulab 2+Uref 2)1/2 described in KS Q ISO 13528, where x, Ulab, X, and Uref are the test results, the uncertainty of the result, and the certified value and the uncertainty of the value. And if the absolute value |En| is less than 1, it can be evaluated as a satisfied value. As a result of ICP-OES analysis, each concentration of the elements to be measured was almost similar to the certified concentration of the reference material, and the uncertainty was slightly different. Also since evaluation on multi-elements determination had an En-score within 1, it was confirmed that the analysis results satisfied En evaluation.

The massive amount of radioactive waste will generated during decommissioning of nuclear. Among the radioactive waste from these disposal process, 50-55 million tons of concrete waste are included. For safe disposal, it is very important to accurately analyze the concentration of radionuclides, especially 129I and 131I, contaminated concrete. 129I, a long-lived radioisotope of iodine (t1/2=1.57 × 107 y), and 131I (t1/2=8.04 d) are generated from the fission of uranium in nuclear reactors. In Korea, according to the Nuclear Safety and Security Commission (NSSC) radioactive clearance level guide, the limit for radioactive clearance level of 129I is less than 0.01 (Bq/g). Iodine can be absorbed, accumulate in organisms, and exhibit low energy emission compared with cesium, and cobalt. Therefore, it is essential to an accurately separate and analyze iodine radioactive waste. In this study, we focused on the determination of iodine in simulated cement waste form containing KI for the recovery of iodine. We performed cement waste form sieved through a different particle size (0.5 mm < ɸ < 6.35 mm). For the separation of iodine from solid samples with low iodine content, such as soil, sediment, and cement, for sample decomposition associated with solvent extraction using CHCl3 for separation of iodine from the matrix. The separation of iodine in cement waste particles was therefore carried out using an acid leaching method using KI containing cement particles. We observed that cement particle size decreased at 6.35 mm to 0.5 mm with iodine yield decrease at 0.840±0.011 to 0.582±0.010. Thus, in this study, the acid leaching method enables to determination Iodine in cement.