• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.37-43
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• 2012

Radioactive waste transport package was developed to transport eight drums of low and intermediate level waste(LILW) in accordance with the IAEA and domestic related regulations. The package is classified with industrial package IP-2. IP-2 package is required to undergo a free drop test and a stacking test. After free drop and stacking tests, it should prevent the loss or dispersal of radioactive contents, and loss of shielding integrity which would result in more than 20 % increase in the radiation level at any external surface of the package. The objective of this study is to establish the safety test method and procedure for stacking test and to prove the structural integrities of the IP-2 package. Stacking test and analysis were performed with a compressive load equal to five times the weight of the package for a period of 24 hours using a full scale model. Strains and displacements were measured at the corner fitting of the package during the stacking test. The measured strains and displacements were compared with the analysis results, and there were good agreements. It is very difficult to measure the deflection at the container base, so the maximum deflection of the container base was calculated by the analysis method. The maximum displacement at the corner fitting and deflection at the container base were less than their allowable values. Dimensions of the test model, thickness of shielding material and bolt torque were measured before and after the stacking test. Throughout the stacking test, it was found that there were no loss or dispersal of radioactive contents and no loss of shielding integrity. Thus, the package was shown to comply with the requirements to maintain structural integrity under the stacking condition.

• 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.

• Microbiology and Biotechnology Letters
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• v.31 no.1
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• pp.83-89
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• 2003

2,4-Dichlorophenoxyacetic acid (2,4-D) as a widely used herbicide has caused serious environmental problems because of its difficult decomposition in nature. We isolated the strain capable of metabolizing 2,4-D as sole carbon and energy source by an enrichment culture technique from the 2,4-D contaminated soil collected at orchard in Gwangju, Korea. This strain was identified tentatively as Aeromonas sp. NOH2. With this strain, we established the response surface methodology (Box-Behnken Design) to optimize the principle parameters for maximizing biodegradation of 2,4-D such as culture pH, temperature, and nutrient concentration in liquid batch culture. The ranges of parameters were obtained from preliminary works done at our laboratory and chosen as 5.5, 6.5, and 7.5 for pH, 25, 30, and $35^{\circ}C$ for temperature, and 5, 20, and 35 g/1 nutrient concentration. Initial concentration of 2,4-D was 500 ppm and nutrient source was tryptic soy broth. The experimental data were significantly fitted to a second order polynomial equation using multiple regression. The most important parameter influencing 2,4-D degradation and biomass production was nutrient concentration. For 2,4-D degradation, the optimum values of pH and temperature, and nutrient concentration were obtained at pH (6.5), temperature (31.8 to $32.1^{\circ}C$), and nutrient concentration (29.6 to 30.1.0 g/1).

• Journal of the Korean Society of Radiology
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• v.14 no.4
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• pp.455-465
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• 2020

The purpose of this study is to evaluate the photoneutron characteristics generated by the linear accelerator target and collimator. The computer simulation design firstly, consisted of a target, a single material target and a composite material target. Secondly, it consisted of a cone beam and a fan beam depending on the type of the collimator. Finally, the material of the fan beam collimator is composed of a single material composed of only lead (Pb) and a composite material collimator composed of tungsten (W) and lead (Pb). The research method calculated the photoneutron production rate and energy spectrum using F2 tally from the surface of a virtual sphere at a distance of 100 cm from the target. As a result, firstly the photoneutron production rate was 20% difference, depending on the target. Secondly, depending on the type of the collimator, there was a 10% difference. Finally, depending on the collimator material, there was a 40% difference. In the photoneutron energy spectrum, the average photoneutron flux tended to be similar to the photoneutron production rate. As a result, it was confirmed that the 9 MeV linear accelerator photoneutron are production increased more by the collimator than by the target, and by the material, not the type of the collimator. Selecting and operating targets and collimator with low photoneutron production will be the most active radiation protection. Therefore, it is considered that this research can be a useful data for introducing and operating and radiation protection of a linear accelerator for container security inspection.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.13 no.4
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• pp.301-313
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• 2015

The rock cavern storage for spent fuel has been assessed to apply in Korea with reviewing the state of the art of the technologies for surface storage and rock cavern storage of spent fuel. The technical feasibility and economic aspects of the rock cavern storage of spent fuel were also analyzed. A considerable area of flat land isolated from the exterior are needed to meet the requirement for the site of the surface storage facilities. It may, however, not be easy to secure such areas in the mountainous region of Korea. Instead, the spent fuel storage facilities constructed in the rock cavern moderate their demands for the suitable site. As a result, the rock cavern storage is a promising alternative for the storage of spent fuel in the aspect of natural and social environments. The rock cavern storage of spent fuel has several advantages compared with the surface storage, and there is no significant difference on the viewpoint of economy between the two alternatives. In addition, no great technical difficulties are present to apply the rock cavern storage technologies to the storage of domestic spent fuel.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.4
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• pp.295-303
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• 2012

Based on the data observed and analyzed on a groundwater flow system in the KURT (KAERI Underground Research Tunnel) site, the transport of radionuclides, which were assumed to be released at the supposed position, was calculated on the time-domain. A groundwater pathway from the release position to the surface was identified by simulating the groundwater flow model with the hydrogeological characteristics measured from the field tests in the KURT site. The elapsed time when the radionuclides moved through the pathway is evaluated using TDRW (Time Domain Random Walk) method for simulating the transport on the time-domain. Some retention mechanisms, such as radioactive decay, equilibrium sorption, and matrix diffusion, as well as the advection-dispersion were selected as the factors to influence on the elapsed time. From the simulation results, the effects of the sorption and matrix diffusion, determined by the properties of the radionuclides and underground media, on the transport of the radionuclides were analyzed and a decay chain of the radionuclides was also examined. The radionuclide ratio of the mass discharge into the surface environment to the mass released from the supposed repository did not exceed $10^{-3}$, and it decreased when the matrix diffusion were considered. The method used in this study could be used in preparing the data on radionuclide transport for a safety assessment of a geological disposal facility because the method could evaluate the travel time of the radionuclides considering the transport retention mechanism.

• Journal of Information Management
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• v.35 no.1
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• pp.1-27
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• 2004

To share knowledge efficiently among our members on the basis of knowledge management system, first of all, we need to systematically design the knowledge classification scheme that we can classify these knowledge well. The objective of this study is to construct the most suitable knowledge classification scheme that all of us can share them in Korea Atomic Energy Research Institute (KAERI). To construct the knowledge classification scheme all over the our organization, we established a few principles to design it and examined related many classification schemes. And I carried out 3 steps to complete the best desirable KAERI's knowledge classification scheme, that is, (1) the step to design a draft of the knowledge classification scheme, (2) the step to revise a draft of the knowledge classification scheme, (3) the step to verify the revised scheme and to decide its scheme. The scheme completed as a results of this study is consisted of total 218 items : sections of 8 items, classes of 43 items and sub-classes of 167 items. I expect that the knowledge classification scheme designed as the results of this study can be played an important role as the frame to efficiently share knowledge among our members when we introduce knowledge management system in our organization. In addition, I expect that steps to design its scheme as well as this scheme itself can be applied when design a knowledge classification scheme at the other R&D institutes and enterprises.

• Geophysics and Geophysical Exploration
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• v.15 no.4
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• pp.227-234
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• 2012

In this study, we make simulation of density log using a Monte Carlo N-Particle (MCNP) algorithm to make an analysis on density logging under different borehole environments, since density logging is affected by various borehole conditions like borehole size, density of borehole fluid, thickness and type of casing, and so on. MCNP algorithm has been widely used for simulation of problems of nuclear particle transportation. In the simulation, we consider the specific configuration of a tool (Robertson Geologging Co. Ltd) that Korea institute of geoscience and mineral resources (KIGAM) has used. In order to measure accurate bulk density of a formation, it is essential to make a calibration and correction chart for the tool under considerations. Through numerical simulation, this study makes calibration plot of the density tool in material with several known bulk densities and with boreholes of several different diameters. In order to make correction charts for the density logging, we simulate and analyze measurements of density logging under different borehole conditions by considering borehole size, density of borehole fluid, and presence of casing.

• Korean Journal of Materials Research
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• v.11 no.5
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• pp.405-412
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• 2001

Corrosion tests were carried out in $360^{\circ}C$ water and $360^{\circ}C$ 70ppm LiOH solution to investigate the corrosion behavior of new zirconium alloys (Zr-0.4Nb-0.8Sn-xFeCrMn, Zr-0.2Nb-1.1Sn-xFeCrMn, Zr-1.0Nb-xFeCu). Microstructures of tested alloys were analyzed by optical microscope and TEM. The cross-sectional surface and crystalline structure of the oxide layer were analyzed by SEM and XRD. From the results of corrosion test, all the alloys showed higher corrosion rates in $360^{\circ}C$ 70ppm LiOH aqueous solution thats in $360^{\circ}C$ water. Especially, high Nb-containing alloy exhibited the acceleration of corrosion rate in LiOH solution. The low Nb- and Sn-added alloys showed better corrosion resistance than the Sn- free high Nb alloy. from the effect of final annealing on the corrosion, it was observed that the partially recrystallized alloys showed better corrosion resistance than fully recrystallized alloys. This would be related to the size and distribution of the second phase particles.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.12 no.1
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• pp.19-35
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• 2014

A parametric sensitivity to the annual exposure dose rate to the farming exposure group has been probabilistically carried out for three principal elements associated with the nuclide transport behavior in the near-field of the pyroprocessed waste repository system. Credit time for both metal and ceramic containers, annual nuclide release rete, and the degree of loss of bentonite buffer around the container are selected as the elements and investigated for important nuclides. All the elements are shown to be sensitive to the results. Methodology studied through this study and the results are expected to make a good feedback to the repository design. As a follow-up study, separated in Part 2, the A-KRS will be deterministically assessed and then compared among each other with the normal, the worst, and the best case scenarios associated with their extreme values these elements could have.