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Home > Publications > International Concrete Abstracts Portal
Showing 1-5 of 887 Abstracts search results
Document:
CI4703Elkhouly
Date:
March 1, 2025
Author(s):
Ahmed F. Elkhouly, Mohammed H. Hedia, and George Morcous
Publication:
Concrete International
Volume:
47
Issue:
3
Abstract:
Hollow-core (HC) slabs are precast/prestressed concrete members with continuous voids oriented parallel to the span of the slabs. Openings in HC slabs are one of the most common challenges encountered during construction. This study experimentally evaluated the effectiveness of proposed strengthening methods for HC slabs with unforeseen structural openings.
CI4703Bischoff
Peter H. Bischoff
This is the second article in a five-part series on calculating deflections for members not meeting minimum thickness requirements in accordance with ACI CODE-318-19. It reviews the development of a new expression for the effective moment of inertia Ie used to calculate immediate deflection of cracked reinforced (nonprestressed) concrete flexural members at service load and justifies the need for a change.
CI4702Bischoff
February 1, 2025
2
The is the first article in a five-part series on calculating deflections for members not meeting minimum thickness requirements in accordance with ACI CODE-318-19. This article reviews the procedure and uncertainty for computing deflection of reinforced (nonprestressed) concrete flexural members.
SP364_01
December 1, 2024
Deuckhang Lee, Hyo-Eun Joo, Sun-Jin Han, Jae Hyun Kim, and Kang Su Kim
Symposium Papers
364
In current ACI 318 code, crack control design criterion for prestressed concrete (PSC) members is stricter than conventional reinforced concrete (RC) members. In particular, it is stipulated that the net tensile stress of prestressing strands should be controlled under 250 MPa (36.3 ksi) in the serviceability design of PSC members belonging to the Class C section that is expected to be cracked under service load conditions. To this end, the nonlinear cracked section analysis is essentially required to estimate the tensile stress of the prestressing strands under the service loads, which requires cumbersome iterative calculations in practice. This study aims to propose a simplified method to estimate the net tensile stress of the prestressing strands (Δfps) under the service load conditions and also a tabulated checking method whether the net tensile stress (Δfps) exceeds the stress limit with respect to the magnitude of effective prestress (Δfse). Finally, applicability of 2,400 MPa (348 ksi) Grade strands is also experimentally investigated.
DOI:
10.14359/51745453
SP364_3
Minkyu Kim, Tae-Hyun Kwon, Gyeonghee An, and Habeun Choi
The containment structure of a nuclear power plant is the last barrier of defense to maintain safety in the event of a severe accident, and the integrity of the containment building is the last line of defense against the release of radioactive material. Nuclear power plant containment buildings are most commonly constructed of prestressed concrete, but there are also some constructed of steel. In the case of PS concrete containment building, in order to prepare for the increase in internal pressure in the event of a severe accident, compression force is applied using a tendon in advance to secure sufficient safety, but due to the characteristics of concrete, cracks may occur, and these cracks may become a pathway for external leakage of radioactive materials in the event of a severe accident. In addition, a number of corroded cavities and degradation of liner plates have been found in recent Korean nuclear power plants. Therefore, a study to evaluate the safety of PS concrete containment buildings began in 2022, started by the Korea Atomic Energy Research Institute, and will be conducted for eight years until 2029. The purpose of the research can be categorized into two main areas. The first is to derive the probability of failure of concrete containment buildings due to an increase in internal pressure in the event of a severe accident. The second objective is to estimate the amount of radioactive material leakage through cracks in the containment building when cracks occur. The current methods for calculating the amount of leakage are approximate and based on many assumptions, and therefore contain too much uncertainty. The results of this study will be used to determine the probability of damage to the containment building in the event of a severe accident at a nuclear power plant, and to quantitatively evaluate the amount of radioactive material leakage to the outside, thereby quantitatively evaluating the amount of external exposure. This paper describes progress to date and potential outcomes rather than highly technical results.
10.14359/51745455
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