• Proceedings of the Korean Radioactive Waste Society Conference
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• 2004.02a
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• pp.140-142
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• 2004

Since the April of 1978, Korea has strongly relied on the nuclear energy for electricity generation. As of today, eighteen nuclear power plants are in operation and ten are to be inaugurated by 2015. The installed nuclear capacity is 15, 716 MW as of the end of 2002, representing 29.3% of the nation's total installed capacity. The nuclear share in electricity remains around 38.9 at the end of 2002, reaching at the level of 119 billion kWh's. New power reactors, KSNP's (Korea Standard Nuclear Power Plant) are fully based on the domestic technologies. More advanced reactors such as KNGR (Korea Next Generation Reactor) will be commercialized soon. Even though the front end nuclear cycle enjoys one of the best positions in the world, there have been some chronical problems in the back end fuel cycle. That's the one of the reason why we need more active R&D programs in Korea and active international and regional cooperation in this area. The everlasting NIMBY problem hinders the implementation of the nation's radioactive waste management program. We expect that the storage capacity for the LILW(Low and Intermediate Level radioactive Waste) will be dried out soon. The situation for the spent fuel storage is also not so favorable too. The storage pools for spent fuel are being filled rapidly so that in 2008, some AR pools cannot accommodate any more new spent nuclear fuels. The Korean Government in strong association with utilities and national academic and R&D institutes have tried its best effort to secure the site for a LILW repository and a AFR site. Finally, one local community, Buan in Jeonbook Province, submitted the petition for the site. At the end of the last July, the Government announced that the Wido, a small island in Buan, is suitable for the national complex site. The special force team headed by Dr IS Chang, president of KAERI teamed with Government officials and many prominent scholars and journalists agreed that by the evidences from the preliminary site investigation, they could not find any reason for rejecting the local community's offer.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.161-170
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• 2012

To provide domestic values of input parameters in a safety assessment of radioactive waste disposal under domestic deep underground environments, various kinds of experiments have been carried out under KURT (KAERI Underground Research Tunnel) conditions. The input parameters were classified, and some of them were selected for this study by the criteria of importance. The domestic experimental data under KURT environments were given top priority in the data review process. Foreign data under similar conditions to KURT were also gathered. The collected data were arranged and the statistical calculations were processed. The properties and distribution of the data were explained and compared to foreign values in view of their validity. The following parameters were analysed: failure time and early time failure rate of a container, solubility of nuclides, porosity and density of the buffer, and distribution coefficients of nuclides in the geomedia, hydraulic conductivity, diffusion depth of nuclides, groundwater flow rate, fracture aperture, length of internal fracture, and width of faulted rock mass in the host rock.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.151-159
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• 2012

The potential repository domains for A-KRS (Advanced Korean Reference Disposal System for High Level Wastes) in geological characteristics of KURT (KAERI Underground Research Tunnel) facility site were proposed to develop a repository system design and to perform the safety assessment. The host rock of KURT facility site is one of major Mesozoic plutonic rocks in Korean peninsula, two-mica granite, which was influenced by hydrothermal alteration. The topographical features control the flow lines of surface and groundwater toward south-easterly and all waters discharge to Geum River. Fracture zones distributed in study site are classified into order 2 magnitude and their dominant orientations are N-S and E-W strike. From the geological features and fracture zones, the potential repository domains for A-KRS were determined spatially based on the following conditions: (1) fracture zone must not cross the repository; and (2) the repository must stay away from the fracture zones greater than 50 m. The western region of the fracture zones in the N-S direction with a depth below 200 m from the surface was sufficient for A-KRS repository. Because most of the fracture zones in N-S direction were inclined toward the east, we expected to find a homogeneous rock mass in the western region rather than in the eastern region. The lower left domain of potential domains has more suitable geological and hydrogeological conditions for A-KRS repository.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.3
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• pp.143-149
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• 2012

Groundwater flow simulations were performed to obtain data of groundwater flow used in a safety assessment for a hypothetical geological disposal facility assumed to be located in the KURT (KAERI Underground Research Tunnel) site. A regional scale modeling of the groundwater flow system was carried out to make boundary conditions for a local scale modeling. And, fracture zones identified at the study site were involved in the local scale groundwater flow model. From the results of the local scale modeling, a hydraulic head distribution was indicated and it was used in a particle tracking simulation for searching pathway of groundwater from the location of the hypothetical disposal facility to the surface where the groundwater reached. The flow distance and discharge rate of the groundwater in the KURT site were calculated. It was thought that the modeling methods used in this study was available to prepare the data of groundwater flow in a safety assessment for a geological disposal facility of radioactive wastes.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.10 no.1
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• pp.13-19
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• 2012

The long-lived fission products $^{79}Se$ and $^{99}Tc$ have been considered as the major concern nuclides for the disposal of radioactive waste because of their high solubilities and the existence of anionic species in natural water. In this study, the solubilities of $FeSe_2(s)$ and $TcO_2(s)$, known as respective Solubility Limiting Solid Phase (SLSP) of selenium and technetium, were measured in the KURT (KAERI Underground Research Tunnel) groundwater under various pH and redox conditions. And their solubilities and major species were also calculated using geochemical codes under conditions similar to experimental solutions. Experimental results and calculation for $FeSe_2$ show that the solubility of selenium was found to be below $1{\times}10^{-6}mol/L$ under the condition of pH 8~9.5 and Eh=-0.3~-0.4 V while the dominant species was identified as $HSe^-$. For $TcO_2$, the solubility of technetium was found to be $5{\times}10^{-8}{\sim}1{\times}10^{-9}mol/L$ in the solutions of pH 6~9.5 and Eh<-0.1 V, while the dominant species was $TcO(OH)_2$. However, when the Eh of the solution is -0.35 V, $TcO(OH)_3^-$ and $TcO_4^-$ are calculated as the dominant species at pH 10.5~12 and pH>12, respectively.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.11 no.4
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• pp.271-280
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• 2013

Since the decommissioning of nuclear plants and facilities, large quantities of slightly contaminated concrete waste have been generated. In Korea, the decontamination and decommissioning of the KRR-1, 2 at the KAERI have been under way. And concrete waste was generated about 800 drums of 200 L. The conditioning of concrete waste is needed for final disposal. The concrete waste is conditioned as follows: mortar using coarse and fine aggregates is filled void space after concrete rubble pre-placement into 200 L drum. Thus, this research has developed an optimizing mixing ratio of concrete waste, water, and cement and has evaluated characteristics of a cement waste form to meet the requirements specified in disposal site specific waste acceptance criteria. The results obtained from compressive strength test, leaching test, thermal cycling test of cement waste forms conclude that the concrete waste, water, and cement have been suggested to have 75:15:10wt% as the optimized mixing ratio. Also, the compressive strength of cement waste form was satisfied that including fine powder up to maximum 40wt% in concrete debris wastes about 75%. As a result of scale-up test, the mixture of concrete waste, water, and cement is 75:10:15wt% meet the satisfied compressive strength because the free water increased with and increased in particle size.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.2
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• pp.187-195
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• 2006

The regulation for the low and intermediate level radioactive waste to be transferred to the disposal facility, recently revised, require that radioactive waste generators should set up waste certification program to verify the radioactive waste conform to the waste acceptance criteria(WAC) before disposal. The radioactive waste disposal facility, scheduled to be constructed in Korea, will institute WAC for the wastes to be transferred to the facility. This WAC is expected to compose of the requirements for the radiological characterization, physical and chemical characterization, physical/chemical restriction, prohibited item, packaging, identification, labeling, and documentation. For the compliance with this regulation, The radioactive waste generators should verify that the waste meet WAC through performance of the waste certification program and are responsible for handing in all the certification documents to the disposal facility. This waste certification program plan was set up as a preliminary program for the certification of radioactive waste generated in Korea Atomic Energy Research Institute (KAERI) and should be further revised until preparation of WAC by disposal agent.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.321-332
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• 2017

In order to build a general model of a high-throughput uranium electrorefining process according to the electrode configuration, numerical analysis was conducted using the COMSOL Multiphysics V5.3 electrodeposition module with Ordinary Differential Equation (ODE) interfaces. The generated model was validated by comparing a current density-potential curve according to the distance between the anode and cathode and the electrode array, using a lab-scale (1kg U/day) multi-electrode electrorefiner made by the Korea Atomic Energy Research Institute (KAERI). The operating temperature was $500^{\circ}C$ and LiCl-KCl eutectic with 3.5wt% $UCl_3$ was used for molten salt. The efficiency of the uranium electrorefining apparatus was improved by lowering the cell potential as the distance between the electrodes decreased and the anode/cathode area ratio increased. This approach will be useful for constructing database for safety design of high throughput spent nuclear fuel electrorefiners.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.333-341
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• 2017

The Korea Atomic Energy Research Institute (KAERI) constructed the PyRoprocessing Integrated inactive Demonstration facility (PRIDE) to carry out experiments on engineering scale pyroprocessing. PRIDE has a large processing cell that human workers are not allowed to access, and thus the equipment inside is operated from outside using remote handling systems. It is therefore essential to examine the operability and maintainability of the equipment in view of remote handling systems, and the equipment is thoroughly examined in a mockup cell before it is installed in the processing cell. If the equipment is tested in a virtual mockup rather than in a mockup cell, the development cost can be significantly reduced. The PRIDE 3D simulator was integrated for virtual verification of equipment that will be installed in the processing cell. All remote handling devices in the actual PRIDE were also virtually installed in the PRIDE 3D simulator. The 3D model of the equipment was loaded and located in the exact position in the virtual processing cell. A scenario to replace an actual electrode was implemented in the PRIDE 3D simulator. The design of the equipment and the working procedures in the scenario were successfully evaluated. The results demonstrated that the PRIDE 3D simulator can be used successfully as an alternative to actual mockup testing.

• Journal of Korean Ophthalmic Optics Society
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• v.6 no.2
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• pp.71-75
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• 2001

In this paper, a three-dimensional measuring system of thermoluminescence(TL) spectra based on temperature, wavelength and luminescence intensity was introduced. The system was composed of a spectrometer, temperature control unit for thermal stimulation, photon detector and personal computer for control the entire system. Temperature control was achieved by using feedback to ensure a linear-rise in the sample temperature. Digital multimeter(KEITHLEY 195A) measures the electromotive force of Copper-Constantan thermocouple and then transmits the data to the computer through GPIB card. The computer converts this signal to temperature using electromotive force-temperature table in program, and then control the power supply through the D/A converter. The spectrometer(SPEX 1681) is controlled by CD-2A, which is controlled by the computer through RS-232 communication port. For measuring the luminescence intensity during the heating run, the electrometer(KEITHLEY 617) measures the anode current of photomultiplier tube(HAMAMATSU R928) and transmits the data to computer through the A/D converter. And, we measured and analyzed thermoluminescence of $CaSO_4$ : Dy, P using the system. The measuring range of thermoluminescence spectra was 300K-575K and 300~800 nm, $CaSO_4$ : Dy. P was fabricated by the Yamashita's method in Korea Atomic Energy Research Institute(KAERI) for radiation dosimeter. Thermoluminesce spectra of the $CaSO_4$ : Dy, P consist of two main peak at temperature of $205^{\circ}C$, wavelength 476 nm and 572 nm and with minor ones at 658 nm and 749 nm.