• Proceedings of the Korean Institute of Building Construction Conference
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• 2016.05a
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• pp.7-8
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• 2016

We are developing the technology for applying the Rebar Modularization Method to the Nuclear Power Plant Structures. To achieve this, we had developed the elementary technology for applying this method to Nuclear Power Plant Structures efficiently and performed the Mock-Up Test by using the developed elementary technology. By analysing this test result, we deduced the problems and found solutions to solve them.

• Nuclear Engineering and Technology
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• v.48 no.3
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• pp.820-830
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• 2016

Pipe wall thinning by flow-accelerated corrosion and various types of erosion is a significant and costly damage phenomenon in secondary piping systems of nuclear power plants (NPPs). Most NPPs have management programs to ensure pipe integrity due to wall thinning that includes periodic measurements for pipe wall thicknesses using nondestructive evaluation techniques. Numerous measurements using ultrasonic tests (UTs; one of the nondestructive evaluation technologies) have been performed during scheduled outages in NPPs. Using the thickness measurement data, wall thinning rates of each component are determined conservatively according to several evaluation methods developed by the United States Electric Power Research Institute. However, little is known about the conservativeness or reliability of the evaluation methods because of a lack of understanding of the measurement error. In this study, quantitative models for UT thickness measurement deviations of nuclear pipes and fittings were developed as the first step for establishing an optimized thinning evaluation procedure considering measurement error. In order to understand the characteristics of UT thickness measurement errors of nuclear pipes and fittings, round robin test results, which were obtained by previous researchers under laboratory conditions, were analyzed. Then, based on a large dataset of actual plant data from four NPPs, a quantitative model for UT thickness measurement deviation is proposed for plant conditions.

• Nuclear Engineering and Technology
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• v.48 no.6
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• pp.1404-1411
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• 2016

Pipe wall thinning in the secondary piping system of a nuclear power plant is currently a major problem that typically affects the safety and reliability of the nuclear power plant directly. Regular in-service inspections are carried out to manage the piping system only during the overhaul. Online thickness monitoring is necessary to avoid abrupt breakage due to wall thinning. To this end, a transducer that can withstand a high-temperature environment and should be installed under the insulation layer. We propose a thin plate type of embedded ultrasonic transducer based on magnetostriction. The transducer was designed and fabricated to measure the thickness of a pipe under a high-temperature condition. A number of experimental results confirmed the validity of the present transducer.

• Corrosion Science and Technology
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• v.14 no.1
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• pp.1-11
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• 2015

Local wall thinning and integrity degradation caused by several mechanisms, such as flow accelerated corrosion (FAC), cavitation, flashing and/or liquid drop impingements, are a main concern in carbon steel piping systems of nuclear power plant in terms of safety and operability. Thinned pipe management program (TPMP) had been developed and optimized to reduce the possibility of unplanned shutdown and/or power reduction due to pipe failure caused by wall thinning in the secondary side piping system. This program also consists of several technical elements such as prediction of wear rate for each component, prioritization of components for inspection, thickness measurement, calculation of actual wear and wear rate for each component. Decision making is associated with replacement or continuous service for thinned pipe components. Establishment of long-term strategy based on diagnosis of plant condition regarding overall wall thinning is also essential part of the program. Prediction models of wall thinning caused by FAC had been established for 24 operating nuclear plants. Long term strategies to manage the thinned pipe component were prepared and applied to each unit, which was reflecting plant specific design, operation, and inspection history, so that the structural integrity of piping system can be maintained. An alternative integrity assessment criterion and a computer program for thinned piping items were developed for the first time in the world, which was directly applicable to the secondary piping system of nuclear power plant. The thinned pipe management program is applied to all domestic nuclear power plants as a standard procedure form so that it contributes to preventing an accident caused by FAC.

• Nuclear Engineering and Technology
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• v.56 no.6
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• pp.2352-2366
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• 2024

Reinforced concrete (RC) shear walls with precast steel-concrete composite modular (PSCCM) are strongly recommended in the structural design of nuclear power plants due to the need for a large number of process pipeline crossings and industrial construction. However, the effect of the PSCCM on the mechanical behavior of the whole RC shear wall is still unknown and has received little attention. In this study, three 1:3 scaled specimens, one traditional shear wall specimen (TW) and two shear wall specimens with the PSCCM (PW1, PW2), were designed and investigated under cyclic loadings. The failure mode, hysteretic curve, energy dissipation, stiffness and strength degradations were then comparatively investigated to reveal the effect of the PSCCM. Furthermore, numerical models of the RC shear wall with different PSCCM distributions were analyzed. The results show that the shear wall with the PSCCM has comparable mechanical properties with the traditional shear wall, which can be further improved by adding reinforced concrete constraints on both sides of the shear wall. The accumulated energy dissipation of the PW2 is higher than that of the TW and PW1 by 98.7 % and 60.0 %. The failure of the shear wall with the PSCCM is mainly concentrated in the reinforced concrete wall below the PSCCM, while the PSCCM maintains an elastic working state as a whole. Shear walls with the PSCCM arranged in the high stress zone will have a higher load-bearing capacity and lateral stiffness, but will suffer a higher risk of failure. The PSCCM in the low stress zone is always in an elastic working state.

• KSCE Journal of Civil and Environmental Engineering Research
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• v.33 no.6
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• pp.2181-2193
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• 2013

The shear wall with openings built with reinforced concrete(RC) have been elaborately studied by many researchers, whereas the steel plate concrete(SC) wall structure has not been investigated as much. Recent SC wall structures developed in Korea have been partly applied to nuclear power plant structures, although its design specification or guideline for the SC wall structure with openings has not been completed yet. This study based on the account for the area lost to openings evaluates the effects of opening on the structural capacity of the SC structure within nuclear power plant. The results obtained from the study on the area lost to openings have been compared with experimental and numerical studies.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2012.11a
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• pp.201-202
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• 2012

Recently, High-Strength steel reinforcement has been studied throughout the internal and external. One of the advantages using High-Strength steel reinforcement in construction is the economic effect due to the decreasing of its quantity. Also, another good effect is the increases of workability by reason of reducing the congestion. But, realistically it is not used in nuclear power plant construction site because of the restriction of design standard. The purpose of this report secures the reliability and changes the code through the performance evaluation test of the wall using the high-strength steel reinforcement in nuclear power plant.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.532-548
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• 2021

Nuclear power plants contain several monitoring systems that can identify the in-vessel phenomena of a severe accident (SA). Though a lot of analysis and research is carried out on SA, right from the development of the nuclear industry, not all the possible circumstances are taken into consideration. Therefore, to improve the efficacy of the safety of nuclear power plants, additional analytical studies are needed that can directly monitor severe accident phenomena. This paper presents an interacting multiple model (IMM) based fault detection and diagnosis (FDD) approach for the identification of in-vessel phenomena to provide the accident propagation information using reactor vessel (RV) out-wall temperature distribution during severe accidents in a nuclear power plant. The estimation of wall temperature is treated as a state estimation problem where the time-varying wall temperature is estimated using IMM employing three multiple models for temperature evolution. From the estimated RV out-wall temperature and rate of temperature, the in-vessel phenomena are identified such as core meltdown, corium relocation, reactor vessel damage, reflooding, etc. We tested the proposed method with five different types of SA scenarios and the results show that the proposed method has estimated the outer wall temperature with good accuracy.

• Proceedings of the Korean Institute of Building Construction Conference
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• 2014.11a
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• pp.205-206
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• 2014

It is very important to reduce the construction duration of the Reactor Containment Building (RCB) when considering the more than 50 months on average from concrete placement to completion. The purpose of this study attempts to evaluate the single-stage workability of the system given a change in the height of the setting of RCB exterior wall formwork to be used in nuclear power plant construction. As a result of this study, it is possible height of 3.5m~4m uses formwork when analyzing the construction period and material costs an increase in formwork by concrete lateral pressure, to ensure the workability of the RCB exterior wall formwork. Through this study, I want to provide as basic data for the improvement of workability and RCB shortening the construction period.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2119-2129
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• 2020

Flow-accelerated corrosion (FAC), liquid droplet impingement erosion (LDIE), cavitation and flashing can cause continuous wall-thinning in nuclear secondary pipes. In order to prevent pipe rupture events resulting from the wall-thinning, most NPPs (nuclear power plants) implement their management programs, which include periodic thickness inspection using UT (ultrasonic test). Meanwhile, it is well known in field experiences that the thickness measurement errors (or deviations) are often comparable with the amount of thickness reduction. Because of these errors, it is difficult to estimate wall-thinning exactly whether the significant thinning has occurred in the inspected components or not. In the previous study, the authors presented an approximate estimation procedure as the first step for thickness measurement deviations at each inspected component and the statistical & quantitative characteristics of the measurement deviations using plant experience data. In this study, statistical significance was quantified for the current methods used for wall-thinning determination. Also, the authors proposed new estimation procedures for determining local wall-thinning to overcome the weakness of the current methods, in which the proposed procedure is based on analysis of variance (ANOVA) method using subgrouping of measured thinning values at all measurement grids. The new procedures were also quantified for their statistical significance. As the results, it is confirmed that the new methods have better estimation confidence than the methods having used until now.