• Structural Engineering and Mechanics
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• v.90 no.5
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• pp.493-504
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• 2024

This paper presents the methodology and results for the investigation of the structural safety of 40 aged underground water tanks to support the weight of photovoltaic (PV) systems that were supposed to be placed on their roof reinforced concrete (RC) slabs. The investigation procedure included (1) review of available documents; (2) visual inspection of the roof RC slabs; (3) carrying out a series of nondestructive (ND) tests; and (4) analysis of results. Out of the 40 tanks, eleven failed the visual inspection phase and were discarded from further investigation. The roof RC slabs of the tanks that passed the visual inspection were subjected to a series of ND tests that included infrared thermography, impact echo, ultrasonic pulse velocity (UPV), Schmidt hammer, concrete core compressive strength, and water-soluble chloride content. The NDT results proved that eight more tanks were not suitable to support the PV systems. Based on the results of the visual inspection and testing, a probabilistic decision-making criterion was established to reach a decision regarding the structural integrity of the roof slabs. The study concluded that the condition of the drainage filter was essential in protecting the tanks and its intact presence can be used as a strong indication of the structural integrity of the roof RC slabs.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.25 no.7
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• pp.1139-1146
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• 2001

The reactor pressure vessel(RPV) is usually cladded with stainless steel to prevent corrosion and radiation embrittlement, and a number of subclad cracks have been found during an in-service-inspection. These subclad cracks should be assured for a safe operation under normal conditions and faulted conditions such as pressurized thermal shock(PTS). Currently available integrity assessment procedure for an RPV, ASME Code Sec. XI, are built on the basis of linear fracture mechanics (LEFM). In PTS condition, however, thermal stress and mechanical stress give rise to high tensile stress at the cladding and elastic-plastic behavior is expected in this area. Therfore, ASME Code Sec. XI is overly conservative in assessing the structural integrity under PTS condition. In this paper, the fracture parameter (stress intensity factor, K, and RT(sub)NDT) from elastic analysis using ASME Sec. XI and finite element method were validated against 3-D elastic-plastic finite element analyses. The difference between elastic and elastic-plastic analysis became significant with increasing crack depth. Therfore, it is recommended to perform elastic-plastic analysis for the accurate assessment of subclad cracks under TPS which causes plastic deformation at the cladding.

• Journal of the Society of Naval Architects of Korea
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• v.49 no.6
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• pp.447-460
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• 2012

In this study, the shutdown system of the fuel gas supply system is designed based on the Safety Integrity Level of IEC 61508 and IEC 61511. First of all, the individual risk($10^{-4}$/year) and the risk matrix which are the risk acceptance criteria are set up for the qualitative risk assessment such as the HAZOP study. The natural gas leakage at the gas supply pipe is identified as the highest risk among the hazards identified through the HAZOP study and as a safety instrumented function the shutdown function for leakage was defined. SIL 2 and PFD($2.5{\cdot}10^{-3}$) for the shutdown function are determined by the layer of protection analysis(LOPA). The shutdown system(SIS) carrying out the shutdown function(SIF) is verified and designed according to qualitative and quantitative requirements of IEC 61508 and IEC 61511. As a result of SIL verification and SIS conceptual design, the shutdown system is composed of two gas detectors voted 1oo2, one programmable logic solver, and two shutdown valve voted 1oo2.

• Microbiology and Biotechnology Letters
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• v.50 no.1
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• pp.51-62
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• 2022

Lactobacillus rhamnosus SKG34 (LrSKG34), a potential probiotic strain, was successfully isolated from Sumbawa Mare's milk. Our previous studies showed that the strain is resistant to gastrointestinal conditions, possesses antioxidant activity, and lowers blood cholesterol levels. Further clarification of the potential probiotic characteristics and safety assessment are necessary. This study aimed to evaluate the adhesion of LrSKG34 to Caco-2 cell monolayers and its effect on mucosal integrity in vitro. We also examined the LrSKG34 safety profile based on antimicrobial susceptibility testing, haemolytic activity determination, Caco-2 cell monolayer translocation evaluation, and in vivo investigation of the effect of LrSKG34 on the physiology, biochemical markers, and histopathological appearance of major organs in an animal model. LrSKG34 attached to Caco-2 cell monolayers and maintained mucosal integrity in vitro. The typical resistance of lactobacilli to ciprofloxacin, gentamicin, vancomycin, trimethoprim-sulfamethoxazole, and metronidazole was confirmed for LrSKG34. No haemolytic activity was observed on blood agar plates, and no LrSKG34 translocation was observed in Caco-2 cell monolayers. Administration of LrSKG34 to Sprague-Dawley rats did not adversely affect body weight. No abnormalities in hematological parameters, serum biochemistry levels, or histopathological structures of major organs were observed in LrSKG34-treated rats. Collectively, the results implicate LrSKG34 as a promising and potentially safe probiotic candidate for further development.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.14 no.7
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• pp.3121-3126
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• 2013

The integrity evaluation of pipes in nuclear power plant are essential for the safety of reactor vessel, and integrity must be assured when flaws are found. Accurate stress intensity analyses and crack growth rate data of surface-cracked components are needed for reliable prediction of their fatigue life and fracture strengths. Fatigue design and life assessment are the essential technologies to design the structures such as pipe, industrial plant equipment and so on. The effect of crack spacing on stress intensity factor K values was studied using three-dimensional finite element method (FEM). For the case of cylinder under internal pressure, a significant increase in K values observed at the deepest point of the surface crack. Also, this paper describes the fatigue analysis for cracked structures submitted to bending loads.

• Journal of Mechanical Science and Technology
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• v.20 no.12
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• pp.2061-2078
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• 2006

The main purpose of this paper is to establish the high temperature structural integrity evaluating procedures for the next generation reactors, which are to be operated at over 500$^{\circ}C$ and for 60 years. To do this, comparison studies of the high temperature structural design codes and assessment procedures such as the ASME-NH (USA), RCC-MR (France), DDS (Japan), and R5 (UK) are carried out in view of the accumulated inelastic strain and the creep-fatigue damage evaluations. Also the application procedures of the ASME-NH rules with the actual thermal and structural analysis results are described in detail. To overcome the complexity and the engineering costs arising from a real application of the ASME-NH rules by hand, all the procedures established in this study such as the time-dependent primary stress limits, total accumulated creep ratcheting strain limits, and the creep-fatigue damage limits are computerized and implemented into the SIE ASME-NH program. Using this program, the selected high temperature structures subjected to two cycle types are evaluated and the parametric studies for the effects of the time step size, primary load, number of cycles, normal temperature for the creep damage evaluations and the effects of the load history on the creep ratcheting strain calculations are investigated.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.5 no.2
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• pp.13-19
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• 2009

As the operating time in nuclear power plants (NPPs) increases, the integrity of nuclear components may be continually degraded due to aging effects of systems, structures and components. Recently, a number of NPPs are being operated beyond their design life to produce more electricity without shutting down. The critical issue in extending a lifetime is to maintain the level of safety during the extended operation period while satisfying the international regulatory standards. Therefore, it is beneficial to build a monitoring system to measure an aging status. In this paper, the Aging Monitor (AM) based on lots of aging database obtained from the operating plants and research results on the aging effects was developed to monitor, manage and evaluate the aging phenomena systematically and effectively in NPPs. The AM for the CANDU is divided into 6 modules: (1) Aging Alarm/Coloring Monitor, (2) Aging Database, (3) Aging Document, (4) Real-time Integrity Monitor, (5) Surveillance and Inspection Management System, and (6) Continued Operation and Periodic Safety Review (PSR) Safety Evaluation. The proposed system is expected to provide the integrity assessment for the major mechanical components of an NPP under concurrent working environments.

• The Journal of Korean Institute of Communications and Information Sciences
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• v.38C no.11
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• pp.972-981
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• 2013

Future aspects of the has turned into a network centric warfare(NCW). Organically combined wired and wireless networks in a variety of cross-of-the-art combat power factor utilization of information and communication technology is a key element of NCW implementation. At used various information in the NCW must be the confidentiality and integrity excellent then quick situation assessment through reliability the real-time processing, which is the core of winning the war. In this paper, NCW is one of the key technologies of the implementation of 128-bit output stream cipher algorithm is proposed. AES-based stream cipher developed by applying modified OFB mode the confidentiality and integrity as well as hardware implementation to the security and real-time processing is superior.

• Journal of the Microelectronics and Packaging Society
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• v.20 no.4
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• pp.75-79
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• 2013

A power distribution network (PDN) is a network that provides connection between the voltage source supply and the power/ground terminals of a microprocessor chip. It consists of a voltage regulator module, a printed circuit board, a package substrate, a microprocessor chip as well as decoupling capacitors. For power integrity analysis, the board and package layouts have to be transformed into an electrical network of resistor, inductor and capacitor components which may be expressed using the S-parameters models. This modeling process generally takes from several hours up to a few days for a complete board or package layout. When the board and package layouts change, they need to be re-extracted and the S-parameters models also need to be re-generated for power integrity assessment. This not only consumes a lot of resources such as time and manpower, the task of PDN modeling is also tedious and mundane. In this paper, a block-based PDN modeling is proposed. Here, the board or package layout is partitioned into sub-blocks and each of them is modeled independently. In the event of a change in power rails routing, only the affected sub-blocks will be reextracted and re-modeled. Simulation results show that the proposed block-based PDN modeling not only can save at least 75% of processing time but it can, at the same time, keep the modeling accuracy on par with the traditional PDN modeling methodology.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.817-824
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• 2012

The most important component of a nuclear power plant is its nuclear reactor. Studies on the integrity of reactors have become an important part regarding the safety of a nuclear power plant. The US Nuclear Regulatory Commission Regulatory Guide (NRC RG) 1.20 presents a Comprehensive Vibration Assessment Program (CVAP) to be used to verify the structural integrity of the Reactor Vessel Internals (RVI) for flow-induced vibration prior to commercial operation. However, there are few published studies related to the RVI CVAP. We classified the Advanced Power Reactor 1400 (APR1400) RVI CVAP as a non-prototype category-2 reactor as part of an independent validation of its design. The aim of this paper is to present the results of structural response analyses of the Upper Guide Structure (UGS) assembly of the APR1400 reactor. These results show that the UGS and the Inner Barrel Assembly (IBA) meet the specified integrity levels of the design acceptance criteria. The vibration and stress analysis results in this paper will be used as basic information to select measurement locations of the vibration and stress for the APR1400 RVI CVAP.