• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.15 no.4
• /
• pp.391-402
• /
• 2017

The Korea Radioactive Waste Agency (KORAD) has developed a concrete cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. During long-term storage of spent nuclear fuel in concrete casks kept in dry conditions, the integrity of the concrete cask and spent nuclear fuel must be maintained. In addition, the radiation dose rate must not exceed the storage facility's design standards. A suitable shielding design for radiation protection must be in place for the dry storage facilities of spent nuclear fuel under normal and accident conditions. Evaluation results show that the appropriate distance to the annual dose rate of 0.25 mSv for ordinary citizens is approximately 230 m. For a $2{\times}10$ arrangement within storage facilities, rollover accidents are assumed to have occurred while transferring one additional storage cask, with the bottom of the cask facing the controlled area boundary. The dose rates of 12.81 and 1.28 mSv were calculated at 100 m and 230 m from the outermost cask in the $2{\times}10$ arrangement. Therefore, a spent nuclear fuel concrete cask and storage facilities maintain radiological safety if the distance to the appropriately assessed controlled area boundary is ensured. In the future, the results of this study will be useful for the design and operation of nuclear power plant on-site storage or intermediate storage facilities based on the spent fuel management strategy.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.11 no.4
• /
• pp.333-349
• /
• 2013

As the capacity of spent nuclear fuel storage pool at reactor sites becomes saturated in ten years, long term dry storage strategy has been recently discussed as an alternative option in Korea. In this study, we reviewed safety-criteria-related research results on spent nuclear fuel performance and integrity under long-term dry storage and proposed the direction and the scope of future domestic research and development. Creep and hydride effect in relation to the embrittlement are known to be the major degradation mechanisms of the spent fuels during the long term dry storage. However, recent research results showed that hydride reorientation and hydride embrittlement are one of the most critical factors to the spent fuel integrity. Accordingly safety criteria of US and Japan for the storage system are basically founded on those mechanisms. However, in Korea, not only in-pile but out-of-pile experimental data have not been generated to understand fuel cladding degradation and to determine the criteria to ensure the safety. In addition, the transient behavior of the spent fuel during transportation also needs to be thoroughly examined. Therefore, various experimental research and development will be required to establish our own safety criteria for future long-term dry storage of domestic spent fuels.

• Nuclear Engineering and Technology
• /
• v.43 no.6
• /
• pp.583-592
• /
• 2011

The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be $1.04{\times}10^{16}$ Bq, $2.09{\times}10^3$ W, $5.31{\times}10^{14}\;m^3$-water, $4.69{\times}10^5$ kg, and $7.38{\times}10^1\;m^3$, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

• Journal of Information Technology Applications and Management
• /
• v.21 no.4
• /
• pp.65-96
• /
• 2014

To have global competitiveness in newly growing industry, good quality R&D's in convergence technology arerequired. Korean government also started to emphasize the importance of convergence technology as a new engine of growth for the future development. Since 2008, worldwide energy crisis and concerns on low carbon green growth made people focus on the convergence between information technology (IT) and energy technology (ET). However, the R&D performance comparison among the convergence technologies is not well explored so far. Therefore, this study uses Korea's patent citation database to measure the R&D performance of convergence technologies. We adopt technology development cycle, technology spillover analysis, and technology diffusion analysis to see the knowledge flow from R&D industry to others. We find that IT_ET convergence technology generally shows higher R&D performance than other convergence technologies. Contrary to public belief, convergence R&D by big companies has relatively low performance especially measures in technology spillover and technology diffusion. This implies that they might concentrate on delicate Fuel Cell Energy/Environment Technology (FEIT) or Nano Environment/Energy Information Technology (NEIT) rather than general energy information technologies. We also find that Korea's chemical industry may play a crucial role for the growth of other convergence technologies.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.15 no.4
• /
• pp.355-367
• /
• 2017

The decommissioning of Kori unit 1, which was permanently shut down in June of 2017, will be the first instance of the dismantling of a commercial nuclear power plant in Korea. The disposal of waste during the dismantling process accounts for a large part of the total decommissioning cost. Therefore, structures consisting of activated and contaminated concrete must be economically and safely dismantled by establishing a proper dismantling strategy. This study focuses on optimized dismantling and disposal scenarios pertaining to a biological shield. Several dismantling cases, regulations and technologies related to waste treatment as these practices pertain to nuclear power plants are analyzed. To minimize the amount of waste from the biological shield dismantling process, an optimized dismantling scenario is presented and disposal alternatives for dismantled concrete waste are proposed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.19 no.4
• /
• pp.459-467
• /
• 2021

In the majority of countries, the upper limit of buffer temperature in a repository is set to below 100℃ due to the possible illitization. This smectite-to-illite transformation is expected to be detrimental to the swelling functions of the buffer. However, if the upper limit is increased while preventing illitization, the disposal density and cost-effectiveness for the repository will dramatically increase. Thus, understanding the characteristics and creating a database related to the buffer under the elevated temperature conditions is crucial. In this study, a strategy to investigate the bentonite found in Korea under the elevated temperatures from a mineral transformation and radionuclides retardation perspective was proposed. Certain long-term hydrothermal reactions generated the bentonite samples that were utilized for the investigation of their mineral transformation and radionuclide retardation characteristics. The bentonite samples are expected to be studied using in-situ synchrotron-based X-Ray Diffraction (XRD) technique to determine the smectite-to-illite transformation. Simultaneously, the 'high-temperature and high-pressure mineral alteration measurement system' based on the Diamond Anvil Cell (DAC) will control and provide the elevated temperature and pressure conditions during the measurements. The kinetic models, including the Huang and Cuadros model, are expected to predict the time and manner in which the illitization will become detrimental to the performance and safety of the repository. The sorption reactions planned for the bentonite samples to evaluate the effects on retardation will provide the information required to expand the current knowledge of repository optimization.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.13 no.2
• /
• pp.155-161
• /
• 2015

Technetium-99m (^99mTc) is an important, short-lived decay product of molybdenum-99 (⁹⁹Mo), and it is considered the backbone of the modern nuclear diagnostic procedures. Since fission of ²³⁵U is the main source of production of ⁹⁹Mo, either highly-enriched uranium (HEU) targets or low-enriched uranium (LEU) targets are irradiated in the research reactors. The use of LEU targets is being promoted by the international community to avoid the proliferation issues linked with the use of HEU. In order to define the waste management strategy at the planning stage of establishment of an LEU based ⁹⁹Mo production facility, the impact of the use of LEU targets on the radioactive waste stream of the ⁹⁹Mo production facility was analyzed. Because the volume of uranium waste is estimated to increase six times, the use of high uranium density targets and the utilization of hot isostatic pressing were recommended to reduce the increased waste volume from the use of LEU based targets.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.3 no.2
• /
• pp.135-148
• /
• 2005

In the stage of conceptual design for the construction and operation of the geologic repository for radioactive wastes, it is important to consider a repository ventilation system which serves the repository working environment, hygiene & safety of the public at large, and will allow safe maintenance like moisture content elimination in repository for the duration of the repositories life, construction/operation/closure, also allowing safe waste transportation and emplacement. This paper describes the possible ventilation system design criteria and requirements for the prospective Korean radioactive waste repositories with emphasis on the underground rock cavity disposal method in the both cases of low & medium-level and high-level wastes. It was found that the most important concept is separate ventilation systems for the construction (development) and waste emplacement (storage) activities. In addition, ventilation network system modeling, natural ventilation, ventilation monitoring systems & real time ventilation simulation, and fire simulation & emergency system in the repository are briefly discussed.

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

• Nuclear Engineering and Technology
• /
• v.55 no.6
• /
• pp.2230-2245
• /
• 2023

KANT (KAIST Advanced Nuclear Tachygraphy) is a PWR core simulator recently developed at Korea Advance Institute of Science and Technology, which solves three-dimensional steady-state and transient multigroup neutron diffusion equations under Cartesian geometries alongside the incorporation of thermal-hydraulics feedback effect for multi-physics calculation. It utilizes the standard Nodal Expansion Method (NEM) accelerated with various Coarse Mesh Finite Difference (CMFD) methods for neutronics calculation. For thermal-hydraulics (TH) calculation, a single-phase flow model and a one-dimensional cylindrical fuel rod heat conduction model are employed. The time-dependent neutronics and TH calculations are numerically solved through an implicit Euler scheme, where a detailed coupling strategy is presented in this paper alongside a description of nodal equivalence, macroscopic depletion, and pin power reconstruction. For validation of the steady, transient, and depletion calculation with pin power reconstruction capacity of KANT, solutions for various benchmark problems are presented. The IAEA 3-D PWR and 4-group KOEBERG problems were considered for the steady-state reactor benchmark problem. For transient calculations, LMW (Lagenbuch, Maurer and Werner) LWR and NEACRP 3-D PWR benchmarks were solved, where the latter problem includes thermal-hydraulics feedback. For macroscopic depletion with pin power reconstruction, a small PWR problem modified with KAIST benchmark model was solved. For validation of the multi-physics analysis capability of KANT concerning large-sized PWRs, the BEAVRS Cycle1 benchmark has been considered. It was found that KANT solutions are accurate and consistent compared to other published works.