• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.224-233
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• 2017

In Korea, attempts have been made to develop high strength concrete for the safety and design life improvement of nuclear power plants. In this study, the cracking potentials of nuclear power plant-high strength concretes (NPP-HSCs) containing industrial by-products with W/B 0.34 and W/B 0.28, which are being reviewed for their application in the construction of containment structures, were evaluated through autogenous shrinkage, unrestrained drying shrinkage, and restrained drying shrinkage experiments. The cracking potentials of the NPP-HSCs with W/B 0.34 and W/B 0.28 were in the order of 0.34FA25 > 0.34FA25BFS25 > 0.34BFS50 > 0.34BFS65SF5 and 0.28FA25SF5 >> 0.28BFS65SF5 > 0.28BFS45SF5 > 0.28 FA20BFS25SF5, respectively. The cracking potentials of the seven mix proportions excluding 0.28FA25SF5 were lower than that of the existing nuclear power plant concrete; thus, the durability of a nuclear power plant against shrinkage cracking could be improved by applying the seven mix proportions with low cracking potentials.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2017.11a
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• pp.137-138
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• 2017

Generally, a lot of reinforcements are used in nuclear power plant concrete structures in order to improve the structural safety, but it may cause several potential problems due to the overcrowded reinforcement, such as the degradation of concrete quality, the construction delay and the increase of construction cost. In order to resolve these problems, structural test researches and code change studies on using high-strength reinforcement (Gr.80) in unclear power plant structures are under way, and there is good progress in code change of ASM BPVC.III.2 and ACI 349. This purpose of this study is to review the code change status ASM BPVC.III.2, ACI 349 under way to use the high-strength reinforcement in nuclear power plant structures. Also I will introduce the design optimization of NPP structures with high-strength reinforcements in order to maximize the effect and minimize the problem when using the high-strength reinforcements in NPP structures.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2069-2087
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• 2023

In nuclear power plant (NPP) accidents, the containment is subject to high temperatures and high internal pressures, which may further trigger serious chain accidents such as core meltdown and hydrogen explosion, resulting in a significantly higher accident level. Therefore, studying the mechanical performance of a containment under high temperature and high internal pressure is relevant to the safety of NPPs. Based on similarity principles, the 1:3.2 scale model of a prestressed concrete containment vessel (PCCV) of a NPP was designed. The loading method, which considers the thermal-pressure coupling conditions, was used. The mechanical response of the PCCV was investigated with a simultaneous increase in internal pressure and temperature, and the failure mechanism of the PCCV under thermal-pressure coupling conditions was revealed.

• Journal of the Korea institute for structural maintenance and inspection
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• v.28 no.4
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• pp.48-54
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• 2024

NPP (Nuclear power plant) structures have been constructed near to the sea shore line for cooling water and exposed to steel corrosion due to chloride attack. Regarding NPP structures built in the UAE, chloride transport may be more rapid than those in the other regions since the temperature near to the coast is high. In this study, concrete samples with 5,000psi (35MPa) design strength grade were manufactured with the materials and mix proportions, which were the same as used in the UAE NPP structures, then chloride diffusion coefficients were evaluated considering temperature and curing age. The compressive strength and the diffusion coefficient were evaluated and analyzed for the samples with 28 and 91 curing days. In addition, chloride diffusion tests for 91-day-cured condition were carried out in the range of 20℃ to 50℃. The activation energy was obtained through converting the temperature slope to a logarithmic function and it was compared with the previous studies. The proposed activation energy can be useful for a reasonable durability design by using actual temperature-dependent chloride diffusion coefficient.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.386-400
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• 2022

Seismic fragility analysis, a part of seismic probabilistic risk assessment (SPRA), is commonly used to establish the relationship between a representative property of earthquakes and the failure probability of a structure, component, or system. Current guidelines on the SPRA of nuclear power plants (NPPs) used worldwide mainly reflect the earthquake characteristics of the western United States. However, different earthquake characteristics may have a significant impact on the seismic fragility of a structure. Given the concern, this study aimed to investigate the effects of earthquake characteristics on the seismic fragility of concrete containments housing the OPR-1000 reactor. Earthquake time histories were created from 30 ground motions (including those of the 2016 Gyeongju earthquake) by spectral matching to the site-specific response spectrum of Hanbit nuclear power plants in South Korea. Fragility curves of the containment structure were determined under the linear response history analysis using a lumped-mass stick model and 30 ground motions, and were compared in terms of earthquake characteristics. The results showed that the median capacity and high confidence of low probability of failure (HCLPF) tended to highly depend on the sustained maximum acceleration (SMA), and increase when using the time histories which have lower SMA compared with the others.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.22 no.2
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• pp.117-128
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• 2024

The 300 concrete silo systems installed and operated at the site of Wolsong nuclear power plant (NPP) have been storing CANDU spent nuclear fuel (SNF) under dry conditions since 1992. The dry storage system must be operated safely until SNF is delivered to an interim storage facility or final repository located outside the NPP in accordance with the SNF management policy of the country. The silo dry storage system consists of a concrete structure, liner steel plate in the inner cavity, and fuel basket. Because the components of the silo system are exposed to high energy radiation owing to the high radioactivity of SNF inside, the effects of irradiation during long-term storage must be analyzed. To this end, material specimens of each component were manufactured and subjected to irradiation and strength tests, and mechanical characteristics before and after irradiation were examined. Notably, the mechanical characteristics of the main components of the silo system were affected by irradiation during the storage of spent fuel. The test results will be used to evaluate the long-term behavior of silo systems in the future.

• Proceedings of the Korea Concrete Institute Conference
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• 1998.10b
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• pp.649-655
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• 1998

In this study, the personal-computer program was developed to predict prestressing losses containment structures of Nuclear Power Plants by concrete creep and shrinkage. This program is constituted of three parts, which are pre-processor, calculation module and post-processor. Input data for his program are : material properties of concrete, rebar, liner and duct, test results of concrete creep and shrinkage, relative humidity, dimension of containment structures, and the number of prestressing tendon related on containment structures. To obtain better results, this program was made to reflect the prestressing losses due to influence that occurred after prestressing each tendon, thus it can predict prestressing losses and allowable prestressing forces of each tendon. As a case study, this program was applied to containment structures of Youngwang 3 & 4 NPP's and analytical result was compared with test results in In-service Inspection of containment structures. From this comparison, it was proved that this program could well predict prestressing losses by concrete creep and shrinkage.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.417-428
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• 2020

Investigations of the commercial aircraft impact effect on nuclear island infrastructures have been drawing extensive attention, and this paper aims to perform the safety assessment of Generation III nuclear power plant (NPP) buildings subjected to typical commercial aircrafts crash. At present Part III, the local damage of the rigid components of aircraft, e.g., engine and landing gear, impacting the steel concrete (SC) structures of NPP containment is mainly discussed. Two typical SC target panels with the thicknesses of 40 mm and 100 mm, as well as the steel cylindrical projectile with a mass of 2.15 kg and a diameter of 80 mm are fabricated. By using a large-caliber air gas gun, both the projectile penetration and perforation test are conducted, in which the striking velocities were ranged from 96 m/s to 157 m/s. The bulging velocity and the maximal deflection of rear steel plate, as well as penetration depth of projectile are derived, and the local deformation and failure modes of SC panels are assessed experimentally. Then, the commercial finite element program LS-DYNA is utilized to perform the numerical simulations, by comparisons with the experimental and simulated projectile impact process and SC panel damage, the numerical algorithm, constitutive models and the corresponding parameters are verified. The present work can provide helpful references for the evaluation of the local impact resistance of NPP buildings against the aircraft engine.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.381-396
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• 2020

Investigations of the commercial aircraft impact effect on nuclear island infrastructures have been drawing extensive attention, and this paper aims to perform the safety assessment of Generation III nuclear power plant (NPP) buildings subjected to typical commercial aircrafts crash. At present Part I, finite element (FE) models establishment and validations for both the aircrafts and NPP buildings are performed. (i) Airbus A320 and A380 aircrafts are selected as the representative medium and large commercial aircrafts, and the corresponding fine FE models including the skin, beam, fuel and etc. are established. By comparing the numerically derived impact force time-histories with the existing published literatures, the rationality of aircrafts models is verified. (ii) Fine FE model of the Chinese Zhejiang Sanao NPP buildings is established, including the detailed structures and reinforcing arrangement of both the containment and auxiliary buildings. (iii) By numerically reproducing the existing 1/7.5 scaled aircraft model impact tests on steel plate reinforced concrete (SC) panels and assessing the impact process and velocity time-history of aircraft model, as well as the damage and the maximum deflection of SC panels, the applicability of the existing three concrete constitutive models (i.e., K&C, Winfrith and CSC) are evaluated and the superiority of Winfrith model for SC panels under deformable missile impact is verified. The present work can provide beneficial reference for the integral aircraft crash analyses and structural damage assessment in the following two parts of this paper.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2012.11a
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• pp.197-198
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• 2012

If the mechanical development be applied to the Nuclear Power Plant Structures instead of the standard hook development, the problem of overcrowding re-bars in the anchorage zone can be solved and the construction quality of the concrete work will be improved. But there are some problems in applying it to the NPP structures because of the restriction on the re-bar yield strength and diameter. After the performance evaluation test for the mechanical development, we can develop the new design equation of the mechanical development length in order to solve the limitation and apply it to NPP structures.