• Proceedings of the Korea Concrete Institute Conference
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• 2004.11a
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• pp.735-738
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• 2004

In this study, it was founded to make the optimal mixture for producing concrete which is self-compacting, yet, and generates low heat of hydration by using fly ash, blast furnace slags and limestone powders as binders in addition to cement while using super-plasticizers and viscosity agents as admixture agents. The structural behaviors of the concrete produced with the selected mixture were compared with those of the concrete currently using for construction of nuclear power plants. The study shows that the blended high fluidity concrete including limestone is better in workability and durability than the concrete currently in use for nuclear power plants.

• Proceedings of the Korea Concrete Institute Conference
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• 2004.11a
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• pp.739-742
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• 2004

In this study, it was founded to make the optimal mixture for producing concrete which is self-compacting, yet, and generates low heat of hydration by using flyash, blast furnace slags and limestone powders as binders in addition to cement while using super-plasticizers and viscosity agents as admixture agents. The structural behaviors of the concrete produced with the selected mixture were compared with those of the concrete currently using for construction of nuclear power plants. The study shows that the blended high fluidity concrete including limestone is better in workability and durability than the concrete currently in use for nuclear power plants.

• Magazine of the Korea Concrete Institute
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• v.7 no.2
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• pp.52-59
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• 1995

• Magazine of the Korea Concrete Institute
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• v.23 no.4
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• pp.28-32
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• 2011

• Magazine of the Korea Concrete Institute
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• v.12 no.2
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• pp.29-33
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• 2000

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.391-402
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• 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)
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• v.21 no.3
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• pp.411-417
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• 2023

During the operation of a nuclear power plant (NPP), the generation of radioactive waste, including dry active waste (DAW), concentrates, spent resin, and filters, mandates the implementation of appropriate disposal methods to adhere to Korea's waste acceptance criteria (WAC). In this context, this study investigates the potential use of polymer concrete (PC) as a high-integrity container (HIC) material for solidifying and packaging these waste materials. PC is a versatile composite material comprising binding polymers, aggregates, and additives, known for its exceptional strength and chemical stability. A comprehensive analysis of PC's long-term integrity was conducted in this study. First, its compressive strength, which is crucial for ensuring the structural stability of HICs over extended periods, was evaluated. Subsequently, the resilience of PC was tested under various stress conditions, including biological, radiological, thermal, and chemical stressors. The findings of this study indicate that PC exhibits remarkable long-term properties, demonstrating exceptional stability even when subjected to diverse stressors. The results therefore underscore the potential viability of PC as a reliable material for constructing high-integrity containers, thus contributing to the safe and sustainable management of radioactive waste in NPPs.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.2866-2877
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• 2021

The complex coupling relationship between liquid sodium and concrete materials affects both the sodium fire characteristics and concrete properties through heat and chemical erosion. In this study, experiments on direct and indirect (separated by a steel plate) contact of the columnar sodium fire with the concrete surface were performed. It was found that the combustion efficiency of liquid sodium in direct contact with concrete was significantly enhanced and accompanied by intermittent explosions and splashing of small concrete fragments. The sodium fire on the surface of the concrete considerably increased the internal temperature, pore size, and distribution density of the concrete. In addition, the depth of influence on the loosening of the concrete structure was also greatly extended. The contact of liquid sodium with the concrete substantially affected its permeability resistance. The water absorption of the concrete surface was increased by more than 70% when liquid sodium directly impacted the bare concrete surface. However, the change in water absorption in the centre of the concrete was primarily affected by the duration of the external heat.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.121-122
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• 2017

• Proceedings of the Korea Concrete Institute Conference
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• 2006.05a
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• pp.42-45
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• 2006

This study investigated headed bar anchorage of exterior beam column joints in nuclear power plants. In nuclear power plant structures, anchorage of headed bar is recommended to satisfy ACI 349-01 App. B that are based on the Concrete Capacity Design (CCD) method. However, CCD method may lead to very conservative results for beam column joints where head is anchored within the diagonal strut and concrete is confined by transverse rebar. Compared with results of 5 joint specimens, the anchorage capacities calculated by ACI 349-01 are underestimated by 70-90%. Therefore, it is necessary to amend ACI 349-01 for the mechanical anchorage in beam column joints.