• KSCE Journal of Civil and Environmental Engineering Research
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• v.35 no.2
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• pp.265-274
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• 2015

The structural safety of nuclear power plant against impact from aircraft crash has been performed so far in two viewpoints such as local behavior and global behavior, and the local behavior has been evaluated using local damage evaluation formulas suggested based on the results of experimental data of RC (Reinforcement Concrete) wall. However, few data have been collected from recent research to evaluate the local behavior and damage of SC (Steel plate reinforced Concrete) wall, which is recently applied to the newly designed nuclear power plant. In this study, local damages of SC wall and RC wall against an idealized aircraft engine projectile impact are evaluated through FE simulation analyses with various wall thicknesses and steel ratio. Through analysis of local collision simulation results of SC and RC wall, the penetration depth of SC wall and RC wall are compared.

• Journal of the Korean Society of Earth Science Education
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• v.6 no.3
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• pp.174-184
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• 2013

The purpose of this study was to investigate the awareness of elementary school teachers on the socio-scientific issues. Fukushima Nuclear Power Plant Accident was used by concrete issue connected with SSI for this study. Participants in this study were twelve elementary school teachers studying at K University Graduate School of Education, located in the central region, who underwent a semi-structured interview. The study method was the phenomenological research method which is one of the qualitative research methods, and the interview papers had been examined by three scientific experts. As a result of the study, it was divided into twenty six themes, eight theme clusters, and two categories, and considered the thoughts on the Fukushima nuclear power plant accident, its influence on Korea, the relationship between science and society as a result of the Fukushima nuclear power plant accident, interested in social issues related to science, application in class, response from students and the influence on students. Teachers had a general understanding of science-related social issues, but did not have much interest in the subject. However, they mentioned that to apply the issues in the curriculum would have a positive influence and encourage scientific motivation in students and, furthermore, helped them to develop the awareness of science in their surroundings. A greater interest in socio-scientific issues need to require from teachers and, through including these issues in the curriculum, we should have positive influence in developing science education.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.410-411
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• 2018

• Magazine of RCR
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• v.16 no.3
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• pp.12-20
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• 2021

• Proceedings of the Korea Concrete Institute Conference
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• 2004.05a
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• pp.208-211
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• 2004

In order to obtain improved concrete mix proportion on nuclear power plant structures, the properties of normal concrete is compared with self-compacting concrete. In addition, various mixes of self-compacting concrete utilizing melaminic acid based admixture is mutually compared and estimated. Because existing normal concrete mixes might occur high temperature in concrete structure, A new multi-component concrete, which declines hydration heat, is demanded. Therefore, in this study, the possibility of manufacturing self-compacting concrete is verified and what influences melaminic acid and various powders have on the properties of self-compacting concrete are investigated.

• Nuclear Engineering and Technology
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• v.44 no.5
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• pp.523-534
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• 2012

The objective of this study was to experimentally investigate the effect of heat curing methods on the temperature history and strength development of slab concrete exposed to $-10^{\circ}C$. The goal was to determine proper heat curing methods for the protection of nuclear power plant structures against early-age frost damage under adverse (cold) conditions. Two types of methods were studied: heat insulation alone and in combination with a heating cable. For heat curing with heat insulation alone, either sawdust or a double layer bubble sheet (2-BS) was applied. For curing with a combination of heat insulation and a heating cable, an embedded heating cable was used with either a sawdust cover, a 2-BS cover, or a quadruple layer bubble sheet (4-BS) cover. Seven different slab specimens with dimensions of $1200{\times}600{\times}200$ mm and a design strength of 27 MPa were fabricated and cured at $-10^{\circ}C$ for 7 d. The application of sawdust and 2-BS allowed the concrete temperature to fall below $0^{\circ}C$ within 40 h after exposure to $-10^{\circ}C$, and then, the temperature dropped to $-10^{\circ}C$ and remained there for 7 d owing to insufficient thermal resistance. However, the combination of a heating cable plus sawdust or 2-BS maintained the concrete temperature around $5^{\circ}C$ for 7 d. Moreover, the combination of the heating cable and 4-BS maintained the concrete temperature around $10^{\circ}C$ for 7 d. This was due to the continuous heat supply from the heating cable and the prevention of heat loss by the 4-BS. For maturity development, which is an index of early-age frost damage, the application of heat insulation materials alone did not allow the concrete to meet the minimum maturity required to protect against early-age frost damage after 7 d, owing to poor thermal resistance. However, the combination of the heating cable and the heat insulating materials allowed the concrete to attain the minimum maturity level after just 3 d. In the case of strength development, the heat insulation materials alone were insufficient to achieve the minimum 7-d strength required to prevent early-age frost damage. However, the combination of a heating cable and heat insulating materials met both the minimum 7-d strength and the 28-d design strength owing to the heat supply and thermal resistance. Therefore, it is believed that by combining a heating cable and 4-BS, concrete exposed to $-10^{\circ}C$ can be effectively protected from early-age frost damage and can attain the required 28-d compressive strength.

• International Journal of Concrete Structures and Materials
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• v.10 no.1
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• pp.47-60
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• 2016

The objective of this study was to evaluate the capability of different strength-maturity models to account for the effect of the hydration heat on the in-place strength development of high-strength concrete specifically developed for nuclear facility structures under various ambient curing temperatures. To simulate the primary containment-vessel of a nuclear reactor, three 1200-mm-thick wall specimens were prepared and stored under isothermal conditions of approximately $5^{\circ}C$ (cold temperature), $20^{\circ}C$ (reference temperature), and $35^{\circ}C$ (hot temperature). The in situ compressive strengths of the mock-up walls were measured using cores drilled from the walls and compared with strengths estimated from various strength-maturity models considering the internal temperature rise owing to the hydration heat. The test results showed the initial apparent activation energies at the hardening phase were approximately 2 times higher than the apparent activation energies until the final setting. The differences between core strengths and field-cured cylinder strengths became more notable at early ages and with the decrease in the ambient curing temperature. The strength-maturity model proposed by Yang provides better reliability in estimating in situ strength of concrete than that of Kim et al. and Pinto and Schindler.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.2
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• pp.91-100
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• 2016

The neutron activation inventories in reactor vessel and its internals, and bio-shield of a PWR nuclear power plant were calculated to evaluate the effect of impurity elements contained in the structural materials on the activation inventory. Carbon steel is, in this work, used as the reactor vessel material, stainless steel as the reactor vessel internals, and ordinary concrete as the bio-shield. For stainless steel and carbon steel, one kind of impurity concentration was employed, and for ordinary concrete five kinds were employed in this study using MCNP5 and FISPACT for the calculation of neutron flux and activation inventory, respectively. As the results, specific activities for the cases with impurity elements were calculated to be more than twice than those for the cases without impurity elements in stainless and carbon steel. Especially, the specific activity for the concrete material with impurity elements was calculated to be 30 times higher than that without impurity. Neutron induced reactions and activation inventories in each material were also investigated, and it is noted that major radioactive nuclide in steel material is Co-60 from cobalt impurity element, and, in concrete material, Co-60 and Eu-152 from cobalt and europium impurity elements, respectively. The results of this study can be used for nuclear decommissioning plan during activation inventory assessment and regulation, and it is expected to be used as a reference in the design phase of nuclear power plant, considering the decommissioning of nuclear power plants or nuclear facilities.

• Journal of the Korean Society of Safety
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• v.23 no.2
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• pp.21-29
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• 2008

To evaluate the structural integrity of the NPP containment building more rigorously, the effective prestress, which is one of the most affecting elements, needs to be estimated exactly. This paper presents the results of an experimental study to determine the effective prestress force in prestressed concrete beams. It is possible to improve the effective prestress measuring method by test beam, which is being applied for the investigation of the nuclear power plant in operation. If experimentally evaluated Lift-Off method in this study can be coupled with test beam test currently being used in in-service nuclear power plant, it is possible to measure prestress loss of the tendon and the level of the effective prestress load.

• Journal of the Korea institute for structural maintenance and inspection
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• v.7 no.2
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• pp.149-157
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• 2003

The purpose of this paper is to quantitatively investigate aging state of epoxy coating on containment structure at nuclear power plant. In order to evaluate an physical bonding of the epoxy coating, adhesion test was performed on a degraded epoxy coating on concrete specimens fabricated by accelerated aging experiment. In addition, impedance data by ultrasonic test were measured to compare with adhesion data. From almost 50 % of the specimens, aging phenomena of epoxy coating such as pin hole, blistering was discovered. To improve reliability on quality degradation of epoxy, co-relation between two kinds of different data was analyzed. By tracing co-related these data, it was possible to figure out physical state of as-built epoxy coating. The possibility to develop new methodology of time - dependent aging state on epoxy coating was found and discussed.