• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.4
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• pp.311-320
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• 2006

Removal characteristics of $H^{14}CO_3$ ion from IRN-150 mixed resin contaminated with $^{14}C$ radionuclide and a gasification behavior of $^{14}C$ radionuclide to $^{14}CO_2$ were investigated. The stripping solutions used for the removal of $^{14}C$ from spent resin were $NaNO_3,\;Na_3PO_4,\;NH_4H_2PO_4,\;H_3PO_4$. The influence of stripping solution concentration on the desorption characteristics of inactive $HCO_3$ ion into stripping solution from IRN-150 mixed resin and the gasification of this ion to $CO_2$ was analyzed. The gasification behavior to $CO_2$ by using NaOH, $HNO_3$, HCl was also compared to that of phosphate solution. Real spent resin stored in Wolsung nuclear power plant was used to evaluate the gasification characteristics of $^{14}C$ radionuclide to $^{14}CO_2$. Gamma radionuclides such as $^{137}Cs,\;^{60}Co$ in residual striping solutions after desorption experiment were analyzed.

• Journal of Radiation Protection and Research
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• v.16 no.2
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• pp.27-39
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• 1991

The diffusion of Sr-85, Cs-137, Co-60 and Am-241 in compacted domestic bentonite was studied, using a diffusion cell unit in which diffusion took place axially from the center of cylindrical bentonite sample body. The effects of compaction density and heat-treated bentonite on diffusion were analysed. And the diffusion mechanism of radionuclide was also analysed by evaluating the measured diffusivity of anion Cl-36. The apparent diffusivities obtained for Sr-85, Cs-137, Co-60 and Am-241 were $l.07{\times}10^{-11},\;6.705{\times}10^{-13},\;l.226{\times}10^{-13}\;and\; l.310{\times}10^{-14}m^2/sec$, respectively. When the as-pressed density of bentonite increased from $1.8\;to\;2.0g/cm^3$, the apparent diffusivity of Cs-137 decreased by quarter. In the case of bentonite heat-treated to $150^{\circ}C$, no significant change in diffusivity was observed, which showed the possibility that the domestic bentonite could be used as a chemical barrier to retard the radionuclide migration at below $150^{\circ}C$. From the calculated pore and surface diffusivity, the surface diffusion due to the concentration gradient of radionuclide sorbed on the solid phase was found to dominate greatly in total transport process.

• Journal of Radiation Protection and Research
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• v.22 no.2
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• pp.111-117
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• 1997

Recently, the conclusion of Comprehensive Test Ban Treaty(CTBT) is globally constructing a network system for nuclear test monitoring. The radionuclide experts of the Conference on Disarmament recommended that the detection of nuclear debris in the atmosphere was an essential factor of nuclear test monitoring and proposed the technical requirements. Based on those requirements, atmospheric radionuclide monitoring system to detect nuclear debris generated from the nuclear explosion test was composed. The system is comprised of high volume air sampler(HVAS), filter paper presser and high purity germanium detector(HPGe). Minimum detectable concentrations(MDCs) of the key nuclides requiring in CTBT monitoring strategies are determined by considering of decay time, counting time and flow rate of the high volume air sampler for the rapid explosion and the optimum measurement condition. The results were selected $10{\pm}$2h, $20{\pm}$2h and $850{\pm}50m^3$/h as parameters, respectively. The relation between the natural air-borne radionuclide concentration of $^{212}Pb$ and MDC were calculated which gave effect in the Compton continuum baseline due to those nuclides in the gamma-ray spectroscopy. These results can be used as an actually tool in the CTBT monitoring strategies.

• Journal of Radiation Protection and Research
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• v.27 no.3
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• pp.165-170
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• 2002

Combined with deposition model onto the ground of radionuclides, the influence of radioactive contamination to agricultural products was analyzed due to wet deposition as well as dry deposition from radioactive air concentration during a nuclear emergency. The previous dynamic food chain model, in which initial input parameter is only radionuclide concentrations on the ground, was improved for the evaluating of radioactive contamination to agricultural products from either radionuclide concentrations in air or radionuclide concentrations on the ground. As the results, in case of deposition onto the ground, wet deposition was more dominant process than thy deposition. While the contamination levels of agricultural products were dependent on the a variety of factors such as radionuclides and rainfall rate. It means that the contamination levels of agricultural products are determined from which is more dominant process between deposition on the ground and interception onto agricultural plants.

• Journal of Radiation Protection and Research
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• v.47 no.3
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• pp.143-151
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• 2022

Background: After the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident, biological alterations in the natural biota, including morphological changes of fir trees in forests surrounding the power plant, have been reported. Focusing on the terminal buds involved in the morphological formation of fir trees, this study developed a method for estimating the absorbed radiation dose rate using radionuclide distribution measurements from tree organs. Materials and Methods: A phantom composed of three-dimensional (3D) tree organs was constructed for the three upper whorls of the fir tree. A terminal bud was evaluated using Monte Carlo simulations for the absorbed dose rate of radionuclides in the tree organs of the whorls. Evaluation of the absorbed dose targeted ¹³¹I, ¹³⁴Cs, and ¹³⁷Cs, the main radionuclides subsequent to the FDNPP accident. The dose contribution from each tree organ was calculated separately using dose coefficients (DC), which express the ratio between the average activity concentration of a radionuclide in each tree organ and the dose rate at the terminal bud. Results and Discussion: The dose estimation indicated that the radionuclides in the terminal bud and bud scale contributed to the absorbed dose rate mainly by beta rays, whereas those in 1-year-old trunk/branches and leaves were contributed by gamma rays. However, the dose contribution from radionuclides in the lower trunk/branches and leaves was negligible. Conclusion: The fir tree model provides organ-specific DC values, which are satisfactory for the practical calculation of the absorbed dose rate of radiation from inside the tree. These calculations are based on the measurement of radionuclide concentrations in tree organs on the 1-year-old leader shoots of fir trees. With the addition of direct gamma ray measurements of the absorbed dose rate from the tree environment, the total absorbed dose rate was estimated in the terminal bud of fir trees in contaminated forests.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.4
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• pp.489-500
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• 2022

The decommissioning of nuclear facilities produces various types of radiologically contaminated waste. In addition, dismantlement activities, including cutting, packing, and clean-up at the facility site, result in secondary radioactive waste such as filters, resin, plastic, and clothing. Determining of the radionuclide content of this waste is an important step for the determination of a suitable management strategy including classification and disposal. In this work, we radiochemically characterized the radionuclide activities of filters used during the decommissioning of Korea Research Reactors (KRRs) 1 and 2. The results indicate that the filter samples contained mainly ³H (500-3,600 Bq·g^-1), ¹⁴C (7.5-29 Bq·g^-1), ⁵⁵Fe (1.1- 7.1 Bq·g^-1), ⁵⁹Ni (0.60-1.0 Bq·g^-1), ⁶⁰Co (0.74-70 Bq·g^-1), ⁶³Ni (0.60-94 Bq·g^-1), ⁹⁰Sr (0.25-5.0 Bq·g^-1), ¹³⁷Cs (0.64-8.7 Bq·g^-1), and ¹⁵²Eu (0.19-2.9) Bq·g^-1. In addition, the gross alpha radioactivity of the samples was measured to be between 0.32-1.1 Bq·g^-1. The radionuclide concentrations were below the concentration limit stated in the low- and intermediatelevel waste acceptance criteria of the Nuclear Safety and Security Commission, and used for the disposal of the KRRs waste drums to a repository site.

• Korean Journal of Soil Science and Fertilizer
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• v.19 no.3
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• pp.251-268
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• 1986

Selecting a site for the safe disposal of radioactive waste requires the evaluation of a wide range of geologic, mineralogic, hydrologic, and physicochemical properties. Although highly diverse, these properties are in fact interrelated. Site requirements are also diverse because they are influenced by the nature of the radionuclides in the waste, for example, their half-lives, specific energy, and chemistry. A fundamental consideration in site selection is the mineralogy of the host rock, and one of the most ubiquitous mineral groups is clay minerals. Clays and clay minerals as in situ lithologic components and engineered barriers may playa significant role in retarding the migration of radionuclides. Their high sorptivity, longevity (stability), low permeability, and other physical factors should make them a very effective retainer of most radionuclides in nuclear wastes. There are, however, some unanswered questions. For example, how will their longevity and physicochemical properties be influenced by such factors as radionuclide concentration, radiation intensity, elevated temperatures, changes in redox condition, pH, and formation fluids for extended periods of time? Understanding of mechanisms affecting clay mineral-radionuclide interactions under prevailing geochemical conditions is important; however, the utilization of experimental geochemical information related to physicochemical properties of clays and clay-bearing materials with geohydrologic models presents a uniquely challenging problem in that many assessments have to be based on model predictions rather than on experiments. These are high-priority research investigations that need to be addressed before complete reliance for disposal area performance is made on clays and clay minerals.

• Journal of Radiation Protection and Research
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• v.21 no.1
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• pp.27-39
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• 1996

The change in spent fuel characteristics by DUPIC fuel cycle(burnup of spent PWR fuel again in CANDU) is examined with time elapse since discharge. Major characteristics examined include isotopic concentration, radioactivity, decay heat radiotoxicity and radiation source-term of spent fuel material, which is existing in a type of spent PWR and DUPIC fuel. Behaviors of major nuclides contributing to such changes are also analyzed in terms of radionuclide concentration. From the analysis, the change in radionuclide concentration by DUPIC shows approximately 2% decrease in actinides concentration and 20% increase in fission products concentration. Radioactivity and decay heat of spent DUPIC fuel does not depend upon radionuclides concentrations, which is a unique in sence of general characteristics of spent fuel. In terms of gamma spectrum, spent DUPIC fuel shows lower values than that of spent PWR fuel by 40 to 50% in the range of $0.01{\sim}0.575$ MeV but much higher over 3.5MeV. Neutron Intensities of both spent fuels are mainly determined by $({\alpha},\;n)$ reaction and spontaneous fission reaction of actinides. Of them, especially, the spontaneous fission reaction Is a major neutron source-term, which causes that neutron intensities of spent DUPIC fuel $having{\sim}3.3$ times higher Cm-244 concentration are ${\sim}4$ times higher than that of spent PWR fuel.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.1
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• pp.65-82
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• 2017

The result of a preliminary safety assessment that was completed by applying the radionuclide inventory calculated on the basis of available data from radioactive waste generation agencies suggested that many difficulties are to be expected with regard to disposal safety and operation. Based on the results of the preliminary safety assessment of the entire disposal system, in this paper, a unit package exceeding the safety goal is selected that occupies a large proportion of radionuclides in intermediate-level radioactive waste. We introduce restrictions on the amount of radioactivity in a way that excludes the high surface dose rate of the package. The radioactivity limit for disposal will be used as the baseline data for establishing the acceptance criteria and the disposal criteria for each disposal facility to meet the safety standards. It is necessary to draw up a comprehensive safety development plan for the Gyeongju waste disposal facility that will contribute to the construction of a Safety Case for the safety optimization of radioactive waste disposal facilities.

• Journal of Radiation Protection and Research
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• v.35 no.4
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• pp.172-177
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• 2010

The radioactive effluents from pressurized heavy water reactors (PHWRs) are relatively larger than those from pressurized water reactors (PWRs). Futhermore, radioactive effluents from PHWRs are released continuously. Thus, the discharge of radioactive effluents is strictly controlled. To do this, radiation detectors are installed at stacks of reactor buildings to monitor the concentration of radioactive effluents in real-time. Derived release limits (DRLs) of annual discharge are also set up for each radionuclide and effluents are rigidly controlled not to exceed those limits. In this paper, the discharge process of radioactive effluents, the standard for establishment of DRL and its methodology, and currents status for PHWRs were reviewed.