• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.1
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• pp.57-61
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• 2016

The Tomographic Gamma Scan (TGS) technique partitions radioactive waste drums into $10{\times}10{\times}16$ voxels and assays both the density and concentration of radioactivity for each voxel thus providing for improved accuracy, when compared to the traditional Non-Destructive Assay(NDA) techniques. It could decrease the degree of precision measurement since there is a trade-off between spatial resolution and precision. This latter drawback is compensated by expanding the Region of Interest (ROI) that differentiates the full energy peaks, which, in turn, results in an optimized degree of precision. The enlarged ROI, however, increases the probability of interference among those nuclides that emit energies in the adjacent spectrum. This study has identified the cause of such interference for the reference nuclide of the TGS technique, $^{137}Cs$ (661.66 keV, half-life 30.5 years), to be $^{110m}Ag$ (657.75 keV, half-life 249.76 days). A new calibration method of determining the optimized ROI was developed, and its effectiveness in accurately characterizing $^{137}Cs$ and eliminating the interference was further ascertained.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11b
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• pp.280-289
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• 2005

A pyrochemical process has been introduced and utilized so that the transmutation of spent PWR fuel in PEACER can produce mainly low and intermediate level waste for near surface disposal. Major radioactive nuclides from PEACER pyroprocessing are composed of TRU and LLFP. In this study, the requirement for the final waste from PEACER is evaluated based on the methodology for establishment of waste acceptance criteria. Also, sensitivity analysis for several input parameters is conducted in order to determine acceptable decontamination factor (DF) and LLFP removal efficiency and to find out input parameter that extremely have an effect on DE As a result of the study, LLFP removal efficiency, especially Sr-90 and Tc-99, is proved to be a major nuclide which contributes to annual dose by human intrusion scenario rather than TRU DF. More than $98.5\%$ of LLFP have to be removed to meet below dose constraint within the DF more than 5.0E+03. Besides, because of the relative short half-life of Sr-90, the increasing of the institutional control period is recommended for most important input parameter to determine DF.

• Journal of radiological science and technology
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• v.43 no.5
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• pp.389-395
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• 2020

The evaluation of radioactivated components of heavy-ion accelerator facilities affects the safety of radiation management and the exposure dose for workers. and this is an important issue when predicting the disposal cost of waste during maintenance and dismantling of accelerator facilities. In this study, the FLUKA code was used to simulate the proton treatment device nozzle and classify the radio-nuclides and total radioactivity generated by each component over a short period of time. The source term was evaluated using NIST reference beam data, and the neutron flux generated for each component was calculated using the evaluated beam data. Radioactive isotopes caused by generated neutrons were compared and evaluated using nuclide information from the International Radiation Protection Association and the Korea Radioisotope association. Most of the nuclides produced form of beta rays and electron capture, and short-lived nuclides dominated. However, In the case of ⁵⁴Mn, which is a radioactive product of iron, the effect of gamma rays should be considered. In the case of tritium generated from a material with a low atomic number, it is considered that handling care should be taken due to its long half-life.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2
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• pp.179-194
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• 2020

A Bayesian approach was introduced to improve the belief of prior distributions of input parameters for the probabilistic safety assessment of radioactive waste repository. A GoldSim-based module was developed using the Markov chain Monte Carlo algorithm and implemented through GSTSPA (GoldSim Total System Performance Assessment), a GoldSim template for generic/site-specific safety assessment of the radioactive repository system. In this study, sequential Bayesian updating of prior distributions was comprehensively explained and used as a basis to conduct a reliable safety assessment of the repository. The prior distribution to three sequential posterior distributions for several selected parameters associated with nuclide transport in the fractured rock medium was updated with assumed likelihood functions. The process was demonstrated through a probabilistic safety assessment of the conceptual repository for illustrative purposes. Through this study, it was shown that insufficient observed data could enhance the belief of prior distributions for input parameter values commonly available, which are usually uncertain. This is particularly applicable for nuclide behavior in and around the repository system, which typically exhibited a long time span and wide modeling domain.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.1
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• pp.83-90
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• 2017

The pyroprocessing of spent nuclear fuel generates LiCl-KCl eutectic waste salt containing radioactive rare earth nuclides. It is necessary to develop a simple process for the treatment of LiCl-KCl eutectic waste in a hot-cell facility. In this study, capture and solidification of a rare earth nuclide (Nd) in LiCl-KCl eutectic salt using an inorganic composite with a $Li_2O-Al_2O_3-SiO_2-B_2O_3$ system was conducted to simplify the existing separation and solidification process of rare earth nuclides in LiCl-KCl eutectic waste salt from the pyroprocessing of spent nuclear fuel. More than 98wt% of Nd in LiCl-KCl eutectic salt was captured when the mass ratio of the composite was 0.67 over $NdCl_3$ in the eutectic salt. The content of $Nd_2O_3$ in the Nd captured-composite reached about 50wt%, and this composite was directly fabricated into a homogeneous and chemical resistant glass waste in a monolithic form. These results will be utilized in designing a process to simplify the existing separation and solidification process.

• Economic and Environmental Geology
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• v.51 no.2
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• pp.177-183
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• 2018

This study assessed an application of electrocoagulation (EC) for the removal of cobalt (Co) in radioactive liquid waste from nuclear power plant. The EC process is an electrochemical means to remove a contaminant in wastewater and a novel process to complement the disadvantage of chemical treatment and membrane process. Radioactive liquid waste has been produced from washing process of radio nuclide power plant cooling system. The EC process eliminates Co from the electrolyte within 10 min; in addition, the dewatered sludge produced in EC process is only 0.2 g. Therefore, the EC process is a promising technique for the removal of Co in radioactive liquid waste and volume reduction of wastes.

• Journal of the Korean Society of Systems Engineering
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• v.17 no.1
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• pp.76-84
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• 2021

Radioactive waste generated from nuclear power plant decommissioning shall be strictly managed so that radioactive materials above the allowable limit are not leaked into the environment. Radioactive wastes shall be classified and treated for management based on characteristics such as the type of waste, physicochemical properties, nuclide concentration and radioactivity. Waste characterization report shall be prepared and submitted to the disposal facility operator to ensure that the treated waste is suitable for disposal. The disposal facility operator shall review the waste Characterization report and visit the nuclear power plant decommissioning site to ensure that the wastes are processed step by step according to the plan. The waste Characterization report may be used as input data to evaluate disposal facility safety. Domestic and foreign data are collected and reviewed to confirm the entire processes from waste generation to delivery. This paper proposes the method to prepare the waste Characterization report which contains data and information on waste characteristics, treatment facilities & method and packaging method & container.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.3
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• pp.265-279
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• 2017

The Gyeongju underground silo type disposal facility, approved for use in December 2014, is in operation for the disposal of low and very low-level radioactive wastes, excluding intermediate-level waste. That is why the existing low-level radioactive waste level has been subdivided and the concentration limit value for intermediate-level waste has been changed in accordance with Nuclear Safety Commission Notice 2014-003. For the safe disposal of intermediate-level wastes, new optimization methodology for calculating the concentration limit of intermediate radioactive level wastes at an underground silo type disposal facility was developed. According to the developed optimization methodology, concentration limits of intermediate-level wastes were derived and the inventory of radioactive nuclides was evaluated. The operation and post closure scenarios were evaluated for the derived radioactive nuclide inventory and the results of all scenarios were confirmed to meet the regulatory limit. However, in case of $^{14}C$, it was confirmed that additional radioactivity limitation through a well scenario was needed in addition to the limit of disposal concentration. It was confirmed that the derived intermediate concentration limit of radioactive waste can be used as the intermediate-level waste concentration limit for the underground disposal facility. For the safe disposal of intermediate-level wastes, KORAD plans to acquire additional data from the radioactive waste generator and manage the cumulative radioactivity of $^{14}C$.

• The Korean Journal of Nuclear Medicine Technology
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• v.16 no.2
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• pp.12-17
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• 2012

Purpose : The patient's clothes and sheet after radioiodine therapy must be disposed of by related regulation. That must be disposed of as radioactive wastes, but that is reusing after radioactivity decay by keeping for the certain period of time. In general, The minimum storage period calculate by standard of take radioactive substance out of radiation controlled area based on measured surface contamination level. But the measurements of surface contamination level are able to differ by measurement method. In this paper, I wish to calculate the minimum storage period of patient's clothes and sheet after radioiodine therapy by measure nuclide concentration offered by the regulation on self-disposal of radioactive wastes. Materials and Methods : The whole area of patient's clothes and sheet measured 31 patients(male:9 patients, female:22 patients), who had radioiodine therapy(3.7 GBq:13 patients, 5.55 GBq:16 patients, 7.4 GBq:2 patients) from july 2011 to march 2012. The minimum storage period is calculated by the regulation on self-disposal of radioactive waste(100 Bq/g) and standard of take radioactive substance out of radiation controlled area(4 kBq/m2) Results : The minimum storage period of pillow sheet, upper uniform, lower uniform by standard of take radioactive substance out of radiation controlled area were each 4.6 days, 63days, 78 days. The minimum storage period of pillow sheet, upper uniform, lower uniform by the regulation on self-disposal of radioactive waste were each 18.1 days, 43 days, 62 days. Conclusion : We can verify that patient's clothes and sheet after radioiodine therapy exists a great deal of radioactive contamination. The minimum storage period calculation of patient's clothes and sheet is better suited to applying nuclide concentration offered by the regulation on self-disposal of radioactive waste. I recommend, To keep for at least 2 months of the patient's clothes and sheet contaminated radioactivity, for prevent contamination and unnecessary radiation exposure.

• Nuclear Engineering and Technology
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• v.25 no.4
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• pp.539-547
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• 1993

A stochastic approach using continuous time Markov process is presented to model the one-dimensional nuclide transport in fractured rock media as a further extension for previous works［1-3］. Nuclide transport of decay chain of arbitrary length in the single planar fractured rock media in the vicinity of the radioactive waste repository is modeled using a continuous time Markov process. While most of analytical solutions for nuclide transport of decay chain deal with the limited length of decay chain, do not consider the case of having rock matrix diffusion, and have very complicated solution form, the present model offers rather a simplified solution in the form of expectance and its variance resulted from a stochastic modeling. As another deterministic way, even numerical models of decay chain transport, in most cases, show very complicated procedure to get the solution and large discrepancy for the exact solution as opposed to the stochastic model developed in this study. To demonstrate the use of the present model and to verify the model by comparing with the deterministic model, a specific illustration was made for the transport of a chain of three member in single fractured rock medium with constant groundwater flow rate in the fracture, which ignores the rock matrix diffusion and shows good capability to model the fractured media around the repository.