Young-Yong Ji

Analytical evaluation of natural radionuclides and their radioactive equilibrium in raw materials and by-products

An investigation into the distribution of natural radionuclides and radioactive secular equilibrium in raw materials and by-products in a domestic distribution was conducted to deduce the optimum conditions for the analytical evaluation of natural radionuclides for (238)U, (226)Ra, and (232)Th using a gamma-ray spectrometer and inductively coupled plasma mass spectrometer (ICP-MS). The range of the specific activities of natural radionuclides was first evaluated by analyzing (228)Ac and (214)Bi, which are (232)Th and (226)Ra indicators, respectively, in about 100 samples of raw materials and by-products through a gamma-ray spectrometer. From further experiments using several samples selected based on the results of the distribution of natural radionuclides, the validation of their analytical evaluations for the indirect measurements using a gamma-ray spectrometer and direct measurements using ICP-MS was assured by comparing their results. Chemically processed products from the raw materials, such as Zr sand and ceramic balls, were generally shown for the type of bead and particularly analyzed showing a definite disequilibrium with above a 50% difference between (238)U and (226)Ra in the uranium series and (232)Th and (228)Ra in the thorium series.

Feasibility about the Direct Measurement of 226 Ra Using the Gamma-Ray Spectrometry

U should be corrected to calculate the net peak area in the energy spec-trum. In general, it is very difficult to conduct peaks stripping with difference of about 0.5 keV, although a HPGe with the superior resolution is applied and the maximum channels is applied to the spectrometer. In this study, several interference correction techniques in the direct measurement were surveyed to evaluate the feasibility for the measurement of

Validation of the quantification of natural radionuclides in raw materials and by-products using gamma-ray spectrometry

The validation of a method for the indirect quantification for 238U, 226Ra, and 232Th activity and the direct quantification for 40K activity using gamma-ray spectrometry was performed in view of consistency, reliability, and accuracy of the results. The gas tightness of Al containers used to confine the radon gas was verified from the establishment of the secular equilibrium between 226Ra and its indicator. To evaluate validation parameters such as linearity, range, and accuracy, it was important to verify the equilibrium state of the reference materials (RM) for U and Th, because the ingrowth of progenies in the uranium decay series can affect the quantification of 226Ra activity even if based on a certified reference material (CRM), while the ingrowth of 228Ra from the thorium decay series should be secured in order to use 228Ac as an indicator of 232Th. In addition, the ruggedness of the method regarding different materials was checked using two kinds of CRM, namely bauxite as an example of a raw material and coal fly ash as a by-product.

Development of the Sampling and Nuclide Analysis Methods for Spent HEPA Filter Wastes

Spent filter wastes of about 2,160 units have been stored in the waste storage facility of the Korea Atomic Energy Research Institute since its operation. These spent filters have generally consisted of a HEPA filter after its filtering of all the contaminants in the gas stream generated during the operation of the HANARO research reactor and the nuclear fuel cycle facilities. At the moment, to secure enough storage space, it is necessary to make a volume reduction of the stored radioactive wastes through a compression treatment or a regulatory clearance. There have been many studies on a treatment and a clearance of the low level radioactive wastes generated from nuclear facilities. These methods are used in view of a reduction of a management cost and disposal cost and the security of free space for a waste storage facility approaching saturation. In order to dispose of the spent filters, it is first necessary to conduct a radionuclide assessment of them. To do that, a sampling procedure should be prepared to obtain a representative sample from a spent filter. As for conducting a nuclide analysis for this representative sample, a corresponding spent filter can be determined as either a regulatory clearance waste or a radioactive waste.Copyright

Simultaneous Determination of the Depth of an Embedded Source and its Radioactivity in the Medium.

AbstractThe simultaneous determination of the depth of an embedded source and its radioactivity in the medium at the environmental surveys is a very useful and advisable method for an in-situ gamma-ray measurement with respect to the time and cost constraint. An algorithm for the determination of the source depth and its radioactivity in the medium was developed using the information on the uncollided photon fluences and measured net count rates, which mean not scattered fluences and background subtracted count rate, at the detector positions. Uncollided photon fluences were calculated at several source depths in the medium as well as at detector positions from the Monte Carlo N-Particle (MCNP) simulation. The results were then used to establish a database to output their values according to the source depth in the medium by inputting the photon energy and detector position from the medium. A simple program about the simultaneous determination of two variables was applied to the results on a task to find out the depth and activity of 134Cs and 137Cs at in-situ gamma-ray spectrometry. Less than 10% and 15% differences compared with the real values at the source depth and radioactivity, respectively, were achieved using the developed program.

Survey Study on Radioactivity of Domestic Fishery Product

Samples of fishery products were tested for radioactivity by using the intake frequency data from Korea Health Statistics. The radioactivity of 40 K, 137 Cs, 134 Cs, and 131 I was analyzed using gamma spectrometry with a simplified sample pre-treatment procedure. The radioactivity range for 40 K was 21.9-3050 Bq/kg, whereas the radioactivities of 137 Cs, 134 Cs, and 131I were under minimum detectable activity which were in the range of 0.140-1.97, 0.0900-1.89 and 0.124-1.94 Bq/kg, respectively, for the three species. The results suggest that the Fukushima accident did not have a significant impact on domestic fishery products, which were analyzed during the period from 2013 to 2015. Additionally, there seemed to be no significant impact of additional exposure dose by the analyzed radionuclides.Samples of fishery products were tested for radioactivity by using the intake frequency data from Korea Health Statistics. The radioactivity of K, Cs, Cs, and I was analyzed using gamma spectrometry with a simplified sample pre-treatment procedure. The radioactivity range for K was 21.9-3050 Bq/kg, whereas the radioactivities of Cs, Cs, and I were under minimum detectable activity which were in the range of 0.140-1.97, 0.0900-1.89 and 0.124-1.94 Bq/ kg, respectively, for the three species. The results suggest that the Fukushima accident did not have a significant impact on domestic fishery products, which were analyzed during the period from 2013 to 2015. Additionally, there seemed to be no significant impact of additional exposure dose by the analyzed radionuclides.

The Spectrometric Determination of the Individual Exposure Rate for Gamma Nuclides From an Environmental Radiation Detector

For making the spectrometric determination of the exposure rate from the environment as well as the radioactive material more practical, an accurate calculation method of the individual exposure rate for the detected gamma nuclides from that spectrum should be suggested without the sophisticated calibration procedure. In this study, the calculation method for the individual exposure rate for detected gamma nuclides from a 3″×3″ NaI(Tl) detector was suggested by introducing the concept of the dose rate spectroscopy and the peak-to-total ratio in the energy spectrum for the exposure rate, which means just a form of multiplied counts and the value of a G-factor in the spectrum.Copyright

Feasibility Study on the Nuclide Analysis of the Radwaste Drum Using the Spectrum to Dose Conversion Factor

Estimating the radioisotope inventory of a drum based on the measured dose rate information, which is called as the dose to curie (DTC) conversion [1–3], has been known that there could be extremely high uncertainty associated with establishing the radioactivity of gamma emitters in a drum. However, the DTC method is still an effective assay method to calculate the radioisotope inventory because of their simple and easy procedures to be applied. To make the DTC method practical, numerous assumptions have to be made and limitations placed on its use. These assumptions and limitations were related to the dose rate measurement and the relative abundance of gamma emitters in a drum. These two variables were generally obtained from the different detection mechanism. Unfortunately, that expanded the limitation of using the DTC method. The spectrum to dose (STD) conversion factor [4,5] that was calculated from the measure pulse height spectrum of the gamma ray detector could be made obtaining two variables from the drum to be assayed at once. This method could be made estimating the radioisotope inventory of a drum more practical.Copyright

Radioactive Analysis of a Spent HEPA Filter Using the Distribution Characteristics of the Captured Radionuclide

To investigate the species and the distribution of the captured nuclides in a HEPA filter, it was dismantled into a metal part and a filter medium part to obtain a filter medium. From the nuclide analysis results for a filter medium part through pre-treatment of it, it was possible to obtain three kinds of typical distributions in view of the distribution of the captured nuclides in the filter medium. When considering these distribution characteristics of the captured radionuclide, the region showing the high concentration of the captured nuclides was the intake or the outlet part of the HEPA filter. On the other side, the middle part generally represented a uniform distribution below the average concentration of the captured nuclides. Therefore, in the event of taking a representative sample of a HEPA filter at the intake and the outlet part, the results of a nuclide analysis for that sample could be estimated as existing in the range from 1.0 to 1.5 times the real concentration of the nuclides captured by a filter medium. As a result, to conduct a radionuclide assessment of a spent HEPA filter without a dismantlement of it, the analysis results for a representative sample taken from both the intake and the outlet part of a spent HEPA filter could be regarded as an average value for the corresponding HEPA filter.

Radiological Dose Assessment for a Regulatory Clearance of the Soil and Concrete Wastes at KAERI

At the KAERI, about 4,800 drums of radioactive contaminated soils and concretes wastes have been stored since their generation in 1988. As these wastes have been stored for more than 18 years, some of them can be regulatory cleared. Before such a regulatory clearance, the radiological characteristics of these wastes should be analyzed first. And based on the analysis result, radiological dose should be estimated to confirm the safety of a disposal of these cleared wastes. In this study, a working procedure for a representative sampling and analysis of the wastes was developed. According to the analysis result, Co-60 and Cs-137 were major radionuclides while some wastes were contaminated with Mn-54, Fe-59, I-131, Cs-134 or Eu-152 of an extremely low concentration. The analysis results show that about 75% of wastes have a radioactivity concentration below 0.1Bq/g and were considered for a clearance. It was assumed that radiation workers are exposed by the inhalation dose and the external dose. The radiological dose was estimated using the RESRAD code for the disposal workers and the management workers as 5.32µSv/yr and 3.13µSv/yr for an individual dose and 1.60E-4 man · Sv/yr and 3.13E-5 man · Sv/yr for a public dose, respectively.