Kun Ho Chung

Design and performance of an automated radionuclide separator: Its application on the determination of 99Tc in groundwater

A modular automated radionuclide separator for (99)Tc (MARS Tc-99) has been developed for the rapid and reproducible separation of technetium in groundwater samples. The control software of MARS Tc-99 was developed in the LabView programming language. An automated radiochemical method for separating (99)Tc was developed and validated by the purification of (99m)Tc tracer solution eluted from a commercial (99)Mo/(99m)Tc generator. The chemical recovery and analytical time for this radiochemical method were found to be 96 ± 2% and 81 min, respectively.

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

Combustion/absorption process for the separation of 14C and 3H in radwastes released from nuclear power plants and their analysis

Radioactivities of 3H and 14C in spent radioactive ion exchange resins and spent radioactive lubricant oils released from nuclear power plants, has been determined using a combustion and sorption method (combustion method). The liquid scintillation counting (LSC) spectra showed that the interference of other radionuclides has not significantly affected the determination of radioactivities of 3H and 14C in the radwaste samples. The chemical structure of 14CO2, which originated from the combustion of radwastes, trapped 14C sorbent has been investigated using Fourier transform infrared spectroscopy (FT-IR). FT-IR study showed interesting results that peaks for uncoupled CO2 and carbonic amide appeared at FT-IR spectra of CO2 high-absorbed 14C sorbents, while the peak for carbamate was only observed at the spectra of CO2 low-absorbed sorbents. During the CO2 sorption in 14C sorbent, temperature and viscosity of the sorbent increased owing to decrease of enthalpy and increase of apparent molecular weight of the sorbent caused by the bonding formation between sorbent molecules.

Analysis and evaluation for consumer goods containing NORM in Korea

We analyzed the consumer goods containing NORM by ICP-MS and evaluated the external dose. To evaluate the external dose, we assumed the small room model as irradiation scenario and calculated the specific effective dose rate using MCNPX code. The external doses for twenty goods are less than 1 mSv considering the specific effective dose rates and usage quantities. However, some of them have relatively high dose and the activity concentration limits are necessary as a screening tool.

Study of the determination of 226 Ra in soil using liquid scintillation counter

This study presented an analytical method for detecting radium in soils using a liquid scintillation counter (LSC). The isotope 226 Ra was extracted from soil using the fusion method and then separated from interfering radionuclides using the precipitation method. Radium was coprecipitated as sulfate salts with barium (Ba) and then converted into Ba(Ra)CO3, which is soluble in an acidic solution. The isotope 222 Rn, the decay progeny of 226 Ra, was trapped in a water immiscible cocktail and analyzed by LSC. The pulse shape analysis (PSA) level was estimated using 90 Sr and 226 Ra standard solutions. The figure of merit was the highest at PSA 80, while the alpha spillover was the lowest at PSA 80. The counting efficiency was 243 ± 2% in a glass vial. This analytical method was verified with International Atomic Energy Agency (IAEA) reference materials, including IAEA-312, IAEA-314, and IAEA-315. The recovery ranged from 60-82%, while the relative bias between the measured value and the recommended value was less than 10%. The minimum detectable activity was 2.1 Bq kg �1 with dry mass

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.

Radiation Dose Assessment of the Uranium Rich Regions in Korea

The radiation dose was assessed for the inhabitants living in the Kumsan, Geosan and Miwon regions which belong to an Okchun metamorphic belt with plenty of uranium resources. The grain, vegetable, milk, and water, which the inhabitants generally ate or drunk, were sampled to reflect the dose effects. And TLD (Thermoluminescence dosimeter) was used for measuring the external dose in the sampled area. An external exposure includes the accumulated gamma dose measured by TLD and an internal exposure includes an ingestion ofthe food, and an inhalation of the air and radon gas. The radionuclides such as 137Cs, 7Be, 40K as well as 238U were included for measuring the radioactivity concentrations of the samples. Radioactivity concentrations for the samples were calculated as averaged annual values. The effective dose by TLD analysis was similar to the domestic average when considering that it was between 0.63 mSv/y and 0.71 mSv/y. The 222Rn concentration of indoor air significantly contributed to the total effective dose where the concentration of these uranium belt regions was more or less higher than the average domestic one. This study revealed that the dose level for the inhabitants ofthe uranium rich regions in Korea was similar to that of the other regions except for the effect by radon.

The Establishment of an In-Situ Real Time Radiation Contour Mapping Technique

The goal of this study is to present that an in-situ radiation dose distribution on an area measured by using a gamma radiation measuring instrument is grasped without a delay. Fundamentally, the test includes an area map as well as real numerical data from the measured area. The map is a computer file with a format of ‘bmp’ which is used in the gamma radiation measuring instrument, and the data is a text file which has numerical values of the latitude, longitude and radiation dose rates of the measured points. The radiation dose rates of the unmeasured points in the area are displayed through an interpolation with those of the measured ones. A 2D plot shows the contour lines with an overlapped map image and a 3D one gives contour lines on the z-axis for an immediate understanding of the radiation level. It is available to optionally display the numerical values of the radiation dose rates and to zoom up and down the image of a certain part for a detailed view. This contour mapping method programmed by using a commercial software requires only a mouse click to show in-situ in a real time the radiation distribution on the corresponding map in a laptop window screen.

The Radioactivity Analysis of 14C in the Graphite Samples From the Dismantled KRR-1&2 Sites by a High Temperature Furnace and a LSC

The radioactivity of 14 C of the graphite samples from the dismantled Korea Research Reactor 1&2 (the KRR-1&2) site was analyzed and proposed to be disposed of as a low level radioactive waste rather than self-disposed of. The graphite wastes, with a weight of seven tons, have been generated during the dismantling of a research reactor with a capacity of one MW from 1995 to 2006. The graphite was used as a moderator for the research reactor and so has been radioactivated by thermal neutron. It was thought that the graphite wastes mainly included a radioisotope of stable carbon, 14 C, a pure beta emitter with a half life of 5,730 years and with a maximum decay energy of 156 keV. Therefore, it has been requested to see whether the dismantled graphite radioactive wastes including 14 C can be self-disposed of or not. In the present study, the radioactivity of 14 C in the graphite sample used in the research reactor was analyzed by using a commercialized high temperature furnace and a Liquid Scintillation Counter (LSC). The combustion temperature of the furnace was five hundred degrees centigrade and especially the temperature in the catalyst region was eight hundred degrees centigrade. The recovery from the furnace was 95% for 14 C and the LSC had a quenching efficiency of approximately 66%. Carbosorb was used as a trapping solution for 14 C. The radioactivity of 14 C was measured by a LSC through the procedure of a pre-treatment such as the combustion of a sample in the temperature range of 500–800 degrees centigrade by a high temperature furnace, trapping of 14 C into carbosorb and cocktailing it with a scintillator. The radioactivity was analyzed to have a concentration with a value of much more than a domestic legal limit for a self-disposal. And an individual effective dose rate estimation was also carried out. Finally, it is suggested that the graphite wastes from the dismantled research reactor should be disposed of at a low level radioactive waste disposal site and monitored.Copyright