• Nuclear Engineering and Technology
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• 제50권6호
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• pp.890-898
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• 2018

Nondestructive inspection (NDI) is an integral part of structural integrity analyses of dry storage casks that house spent nuclear fuel. One significant concern for the structural integrity is stress corrosion cracking in the heat-affected zone of welds in the stainless steel canister that confines the spent fuel. In situ NDI methodology for detection of stress corrosion cracking is investigated, where the inspection uses a delivery robot because of the presence of the harsh environment and geometric constrains inside the cask protecting the canister. Shear horizontal (SH) guided waves that are sensitive to cracks oriented either perpendicular or parallel to the wave vector are used to locate welds and to detect cracks. SH waves are excited and received by electromagnetic acoustic transducers (EMATs) using noncontact ultrasonic transduction and pulse-echo mode. A laboratory-scale canister mock-up is fabricated and inspected using the proposed methodology to evaluate the ability of EMATs to excite and receive SH waves and to locate welds. The EMAT's capability to detect notches from various distances is evaluated on a plate containing 25%-through-thickness surface-breaking notches. Based on the results of the distances at which notch reflections are detectable, NDI coverage for spent nuclear fuel storage canisters is determined.

• 한국정밀공학회:학술대회논문집
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• 한국정밀공학회 2005년도 추계학술대회 논문집
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• pp.808-813
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• 2005

Structural integrity assessment of defected pipe is important in fitness for service evaluation and proper engineering assessment is needed to determine whether pipelines are still fit for service. This paper present a failure prediction of gas pipes made of APIl X65 steel with gouge using stress-modified true fracture strain, which is regarded as a criterion of ductile fracture. For this purpose, API X65 pipes with gouge are simulated using elastic-plastic FE analyses with the proposed ductile failure criterion and the resulting burst pressures are compared with experimental data. Agreements are quite good, which gives confidence in the use of the proposed criteria to defect assessment fer gas pipelines. Then, further extensive finite element analyses are performed to obtain the burst pressure solution of pipes with gouge as a function of defect depth, length and pipeline geometry.

• 한국기계가공학회지
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• 제20권7호
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• pp.1-7
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• 2021

Restrictions on the emissions of nitrogen oxides, sulfur oxides, carbon dioxide, and particulate matter from marine engines are being tightened. Each of these emissions requires different reduction technologies, which are costly and require many pieces of equipment to meet the requirements. Liquefied natural gas (LNG) fuel has a great advantage in reducing harmful emissions emitted from ships. Therefore, the marine engine application of LNG fuel is significantly increasing in new ship buildings. Accordingly, this study analyzed the internal support structure, insulation type, and fuel supply piping system of a 35 m3 International Maritime Organization C type pressurized storage tank of an LNG-fueled ship. Analysis of the heat transfer characteristics revealed that A304L stainless steel has a lower heat flux than A553 nickel steel, but the effect is not significant. The heat flux of pearlite insulation is much lower than that of vacuum insulation. Moreover, the analysis results of the constraint method of the support ring showed no significant difference. A553 steel containing 9% nickel has a higher strength and lower coefficient of thermal expansion than A304L, making it a suitable material for cryogenic containers.

• Wind and Structures
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• 제37권6호
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• pp.461-471
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• 2023

The main objective of this study was to establish design guidelines for three key design variables (spar thickness, spar diameter, and total draft) by examining their impact on the stress distribution and resonant frequency of a 2.5-MW spar-type floating offshore wind turbine substructure under extreme marine conditions, such as during Typhoon Bolaven. The current findings revealed that the substructure experienced maximum stress at wave frequencies of either 0.199 Hz or 0.294 Hz, consistent with previously reported experimental findings. These results indicated that the novel simulation method proposed in this study, which simultaneously combines hydrodynamic diffraction analysis, computational dynamics analysis, and structural analysis, was successfully validated. It also demonstrated that our proposed simulation method precisely quantified the stress distribution of the substructure. The novel findings, which reveal that the maximum stress of the substructure increases with an increase in total draft and a decrease in spar thickness and spar diameter, offer valuable insights for optimizing the design of spar-type floating offshore wind turbine substructures operating in various harsh marine environments.

• 한국산학기술학회논문지
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• 제14권7호
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• pp.3121-3126
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• 2013

원전 배관의 건전성평가는 원자로 안전을 위해 중요하며 결함발견 시 반드시 건전성을 확보해야만 한다. 균열을 갖는 구조물에 대한 정확한 응력확대계수 해석과 균열성장속도는 파괴강도와 피로수명을 평가하는데 필요로 한다. 피로설계와 수명평가는 배관, 산업공장장비 등과 같은 구조물을 설계하는데 극히 중요하다. 응력확대계수를 이용한 균열간의 상호 간섭해석은 유한요소법으로 구하였다. 내압을 받는 원통형구조물의 경우 표면균열의 인접점에서 간섭이 가장 크게 일어남을 확인하였다. 또한, 반복하중 균열에 대해서는 균열 성장평가와 더불어 피로하중에 의한 균열진전을 예측하기 위한 피로해석을 수행하였다.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2004년도 추계학술대회
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• pp.220-225
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• 2004

The mesh-insensitive structural stress procedure by Dong is modified to apply to the welded joints with local thickness variation and inignorable shear/normal stresses along local discontinuity surface. In order to make use of the structural stress based K solution for fatigue correlation of welded joints, a proper crack growth model needs to be developed. There exist some significant discrepancies in inferring the slope or crack growth exponent in the conventional Paris law regime. Two-stage crack growth model was not considered since its applications are focused upon the fatigue behavior in welded joints in which the load ratio effects are considered negligible. In this paper, a two-stage crack growth law considering high mean loading is proposed and proven to be effective in unifying the so-called anomalous short crack growth data.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회A
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• pp.250-255
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• 2007

Steam generator tubes play an important role in safety because they constitute one of the primary barriers between the radioactive and non-radioactive sides of the nuclear power plant. For this reason, the integrity of the tubes is essential in minimizing the leakage possibility of radioactive water. The integrity of the tubes is evaluated based on NDE (non-destructive evaluation) inspection results. Especially ECT (eddy current test) method is usually used for detecting the flaws in steam generator tubes. However, detection capacity of the NDE is not perfect and all of the "real flaws" which actually existing in steam generator tunes is not known by NDE results. Therefore reliability of NDE system is one of the essential parts in assessing the integrity of steam generators. In this study POD (probability of detection) of ECT system for ODSCC in steam generator tubes is evaluated using multivariate logistic regression. The cracked tube specimens are made using the withdrawn steam generator tubes. Therefore the cracks are not artificial but real. Using the multivariate logistic regression method, continuous POD surfaces are evaluated from hit (detection) and miss (no detection) binary data obtained from destructive and non-destructive evaluation of the cracked tubes. Length and depth of cracks are considered in multivariate logistic regression and their effects on detection capacity are evaluated.

• 한국압력기기공학회 논문집
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• 제5권2호
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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.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 추계학술대회
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• pp.1358-1363
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• 2003

Due to the application of Integrated Head Assembly (IHA) in KSNP+ reactor design, an investigation of reactor internals integrity is carried out to assure that the adoption of IHA does not affect the safety of reactor operation. One of the postulated accident events is the R.V. closure head fall from 5.5m high directly above the reactor vessel that may occur during the refueling operation. The analysis model consists of lumped mass elements of the entire reactor vessel and internals. Because of extreme load, separate elastic-plastic analyses are done for the members that undergo plastic deformation. The analysis verified that the stresses of the reactor internals and the fuel assemblies are within the bound of allowable stress limits and the integrity of the fuel assemblies is maintained.

• Nuclear Engineering and Technology
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• 제47권6호
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• pp.766-775
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• 2015

The integrity assessment of reactor vessel internals should be conducted in the design process to secure the safety of nuclear power plants. Various loads such as self-weight, seismic load, flow-induced load, and preload are applied to the internals. Therefore, the American Society of Mechanical Engineers (ASME) Code, Section III, defines the stress limit for reactor vessel internals. The present study focused on structural response analyses of the upper guide structure upper flange. The distributions of the stress intensity in the flange body were analyzed under various design load cases during normal operation. The allowable stress intensities along the expected sections of stress concentration were derived from the results of the finite element analysis for evaluating the structural integrity of the flange design. Furthermore, seismic analyses of the upper flange were performed to identify dynamic behavior with respect to the seismic and impact input. The mode superposition and full transient methods were used to perform time-history analyses, and the displacement at the lower end of the flange was obtained. The effect of the damping ratio on the response of the flange was also evaluated, and the acceleration was obtained. The results of elastic and seismic analyses in this study will be used as basic information to judge whether a flange design meets the acceptance criteria.