• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.4137-4141
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• 2021

As NPPs continue to operate, liquid waste continues to be generated, and containers are needed to store and transport them at low cost and high capacity. To transport and store liquid phase very low-level radioactive waste (VLLW), a container is designed by considering related regulations. The design was constructed based on the existing container design, which easily transports and stores liquid waste. The radiation shielding calculation was performed according to the composition change of barium sulfate (BaSO₄) using the Monte Carlo N-Particle (MCNP) code. High-density polyethylene (HDPE) without mixing the additional BaSO₄, represented the maximum dose of 1.03 mSv/hr (<2 mSv/hr) and 0.048 mSv/hr (<0.1 mSv/hr) at the surface of the inner container and at 2 m away from the surface, respectively, for a 10 Bq/g of ⁶⁰Co source. It was confirmed that the dose from the inner container with the VLLW content satisfied the domestic dose standard both on the surface of the container and 2 m from the surface. Although it satisfies the dose standard without adding BaSO₄, a shielding material, the inner container was designed with BaSO₄ added to increase radiation safety.

• The Journal of Engineering Geology
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• v.19 no.4
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• pp.543-545
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• 2009

• The Journal of Engineering Geology
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• v.14 no.3 s.40
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• pp.325-333
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• 2004

• Nuclear industry
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• v.5 no.11 s.33
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• pp.18-24
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• 1985

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.4
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• pp.439-455
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• 2020

Numerical model was developed that simulates radionuclide (3H and 14C) transport modeling at the 2nd phase facility at the Wolsong LILW Disposal Center. Four scenarios were simulated with different assumptions about the integrity of the components of the barrier system. For the design case, the multi-barrier system was shown to be effective in diverting infiltration water around the vaults containing radioactive waste. Nevertheless, the volatile radionuclide 14C migrates outside the containment system and through the unsaturated zone, driven by gas diffusion. 3H is largely contained within the vaults where it decays, with small amounts being flushed out in the liquid state. Various scenarios were examined in which the integrity of the cover barrier system or that of the concrete were compromised. In the absence of any engineered barriers, 3H is washed out to the water table within the first 20 years. The release of 14C by gas diffusion is suppressed if percolation fluxes through the facility are high after a cover failure. However, the high fluxes lead to advective transport of 14C dissolved in the liquid state. The concrete container is an effective barrier, with approximately the same effectiveness as the cover.

• Journal of Korea Technical Association of The Pulp and Paper Industry
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• v.40 no.3
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• pp.48-52
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• 2008

Waste paper plastic composites were prepared with old newspaper and old corrugated containers and mixed office waste and those properties were evaluated. The results were summarized as fellows. 1. The strength properties like as tensile and Young's modulus reveled most high level in MOW composite. 2. The coagulation of fibers in paper particle should interrupt equal dispersion of polymer and paper particle. 3. The micrograph of the surface of composites showed the most high dispersion in ONP composite.

• Nuclear Engineering and Technology
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• v.32 no.5
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• pp.495-503
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• 2000

The hydraulic conductivities in the bentonite-sand mixtures with high density were measured, and the effects of sand content and dry density on the hydraulic conductivity were investigated. The hydraulic conductivities of the bentonite-sand mixtures with a dry density of 1.6 Mg/㎥ and 1.8 Mg/㎥ are less than 10$^{-11}$ m/s when the sand content is not higher than 70 wt%. However at the sand content of 90 wt%, the hydraulic conductivity increases rapidly At the same dry density, the logarithm of hydraulic conductivity increases linearly with increasing sand content. The hydraulic conductivity of the bentonite-sand mixture can be explained by the concept of effective clay dry density, and using this concept, the hydraulic conductivities for the mixtures with various sand contents and dry densities can be estimated.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4005-4012
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• 2022

As high-level radioactive waste (HLW) generated from nuclear power plants is harmful to the human body, it must be safely disposed of by an engineered barrier system consisting of disposal canisters and buffer and backfill materials. A gap exists between the canister and buffer material in a HLW repository and between the buffer material and natural rock-this gap may reduce the water-blocking ability and heat transfer efficiency of the engineered barrier materials. Herein, the basic characteristics and thermal properties of granular bentonite, a candidate gap-filling material, were investigated, and their effects on the temperature change of the buffer material were analyzed numerically. Heat transfer by air conduction and convection in the gap were considered simultaneously. Moreover, by applying the Korean reference disposal system, changes in the properties of the buffer material were derived, and the basic design of the engineered barrier system was presented according to the gap filling material (GFM). The findings showed that a GFM with high initial thermal conductivity must be filled in the space between the buffer material and rock. Moreover, the target dry density of the buffer material varied according to the initial wet density, specific gravity, and water content values of the GFM.

• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1540-1554
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• 2023

To use nuclear energy sustainably, spent nuclear fuel, classified as high-level radioactive waste and inevitably discharged after electricity generation by nuclear power plants, must be managed safely and isolated from the human environment. In Korea, the land area is limited and the amount of high-level radioactive waste, including spent nuclear fuels to be disposed, is relatively large. Thus, it is particularly necessary to maximize disposal efficiency. In this study, a high-efficiency deep geological repository concept was developed to enhance disposal efficiency. To this end, design strategies and requirements for a high-efficiency deep geological repository system were established, and engineered barrier modules with a disposal canister for pressurized water reactor (PWR)-type and pressurized heavy water reactor type Canada deuterium uranium (CANDU) plants were developed. Thermal and structural stability assessments were conducted for the repository system; it was confirmed that the system was suitable for the established strategies and requirements. In addition, the results of the nuclear safety assessment showed that the radiological safety of the new system met the Korean safety standards for disposal of high-level radioactive waste in terms of radiological dose. To evaluate disposal efficiency in terms of the disposal area, the layout of the developed disposal areas was assessed in terms of thermal limits. The estimated disposal areas were 2.51 km² and 1.82 km² (existing repository system: 4.57 km²) and the excavated host rock volumes were 2.7 Mm³ and 2.0 Mm³ (existing repository system: 4.5 Mm³) for thermal limits of 100 ℃ and 130 ℃, respectively. These results indicated that the area and the excavated volume of the new repository system were reduced by 40-60% compared to the existing repository system. In addition, methods to further improve the efficiency were derived for the disposal area for deep geological disposal of spent nuclear fuel. The results of this study are expected to be useful in establishing a national high-level radioactive waste management policy, and for the design of a commercial deep geological repository system for spent nuclear fuels.

• Journal of the Korean Ceramic Society
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• v.43 no.3 s.286
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• pp.148-151
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• 2006

In order to examine the process parameters for the vitrification of Low and Intermediate Level radioactive Waste (LILW) generated from nuclear power plants, measurements of several melt properties was performed for four selected glasses containing simulated waste. Electrical conductivity and viscosity were determined at temperatures ranging from 1123 to $1673^{\circ}C$. The temperature dependences of both properties in the molten state showed a similar behavior in which their values decrease as the temperature increases. The values of the electrical conductivity and viscosity at a temperature of 1423K adopted in an induction cold crucible melter process were $0.27{\sim}0.42$ S/cm and $9.8{\sim}42$ dPas, respectively.