• Journal of Welding and Joining
• /
• v.34 no.6
• /
• pp.16-19
• /
• 2016

Needs for structural integrity procedure such as BS 7910, the nuclear industry document R6 Rev.4 and the European FITNET procedure are being increased in industry. Especially, BS 7910 allows metallic structures to be assessed on the basis of fracture mechanics analysis rather than strict adherence to design and fabricated codes. This study is to propose the flaw assessment to judge the toughness level of welding consumables at the development stage. The FCA welding consumables with YP 690MPa and CTOD over 0.25 mm have been developed and its allowable weld flaw size considering actually applied environment has been evaluated. Since the estimated allowable defect size is sufficiently detectable in nondestructive testing, the toughness of the developed material is judged to be appropriate and no problem in securing the structural integrity.

• Nuclear Engineering and Technology
• /
• v.45 no.2
• /
• pp.249-256
• /
• 2013

In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful preparation is essential to the measurement program, because data acquisition must be performed only once. The optimized design of a vibration and stress measurement system for the RVI CVAP is essential to verify the integrity of the APR1400 RVI. We successfully designed a vibration and stress measurement system for the APR1400 RVI CVAP based on the design materials, the hydraulic and structural analysis results, and performance tests of transducers in an extreme environment. The measurement system designed in this paper will be utilized for the APR1400 RVI CVAP as part of the first construction project in Korea.

• Proceedings of the Korean Society of Precision Engineering Conference
• /
• 2005.10a
• /
• pp.808-813
• /
• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.21 no.4
• /
• pp.465-479
• /
• 2023

To ensure the safety of disposal facilities for radioactive waste, it is essential to quantitatively evaluate the performance of the waste disposal facilities by using safety assessment models. This paper addresses the development of the safety assessment model for the underground silo of Wolseong Low-and Immediate-Level Waste (LILW) disposal facility in Korea. As the simulated result, the nuclides diffused from the waste were kept inside the silo without the leakage of those while the integrity of the concrete is maintained. After the degradation of concrete, radionuclides migrate in the same direction as the groundwater flow by mainly advection mechanism. The release of radionuclides has a positive linear relationship with a half-life in the range of medium half-life. Additionally, the solidified waste form delays and reduces the migration of radionuclides through the interaction between the nuclides and the solidified medium. Herein, the phenomenon of this delay was implemented with the mass transfer coefficient of the flux node at numerical modeling. The solidification effects, which are delaying and reducing the leakage of nuclides, were maintained the integrity of the nuclides. This effect was decreased by increasing the half-life and the mass transfer coefficient of radionuclides.

• Journal of the Society of Disaster Information
• /
• v.12 no.1
• /
• pp.1-9
• /
• 2016

Local wall thinning elbow due to the water flow is one of the main degradation phenomenon in carbon steel pipe of plant system. The API579 code of FFS (fitness for service) used in the chemical plant gives a guideline to protect local wall thin problem in a straight pipe, but it does not include an elbow. In this study, the locally wall thinned elbow, which is considered an extrados or intrados wall thinned elbow, was carried out considering in-plane or out-plane bending using FEM (finite element method) analysis, and some of results were also compared with the experimental data. The results could give the structure integrity assessment to protect the local wall thinned elbow.

• Nuclear Engineering and Technology
• /
• v.48 no.6
• /
• pp.1423-1432
• /
• 2016

The integrity of a reactor pressure vessel (RPV) related to pressurized thermal shocks (PTSs) has been extensively studied. This paper introduces an integrity assessment of an RPV subjected to a PTS transient based on the French codes. In the USA, the "screening criterion" for maximum allowable embrittlement of RPV material is developed based on the probabilistic fracture mechanics. However, in the French RCC-M and RSE-M codes, which are developed based on the deterministic fracture mechanics, there is no "screening criterion". In this paper, the methodology in the RCC-M and RSE-M codes, which are used for PTS analysis, are firstly discussed. The bases of the French codes are compared with ASME and FAVOR codes. A case study is also presented. The results show that the method in the RCC-M code that accounts for the influence of cladding on the stress intensity factor (SIF) may be nonconservative. The SIF almost doubles if the weld residual stress is considered. The approaches included in the codes differ in many aspects, which may result in significant differences in the assessment results. Therefore, homogenization of the codes in the long time operation of nuclear power plants is needed.

• Nuclear Engineering and Technology
• /
• v.52 no.9
• /
• pp.2092-2099
• /
• 2020

Many damages due to flow-accelerated corrosion and cracking have been observed during recent in-service inspections of nuclear power plants. To determine the operability or repair for damaged pipes, an integrity evaluation related to the damaged piping system should be performed by using already proven code and standards. One of them, the ASME Code Case is most popularly used to integrity assessment in nuclear power plants. However, the recent version of CC N-513 still recommends the simplified method which means a damaged elbow is assumed as an equivalent straight pipe. In addition, to enhance the accuracy integrity assessment in elbow, several previous studies recommend that the SIF and elastic COD values for an elbow with relatively large crack could be predicted by an interpolation technique. However, those estimates for elbow with relatively large crack might be derived to inaccurate results for crack growth analysis, such as for the allowable crack size and life estimation. Therefore, in this paper, the effect of crack length (0.3≤θ₁/π≤0.5) on SIF and elastic COD for elbow is systematically investigated. Then, for large crack in elbow, accurate estimates for SIF and elastic COD, which are widely used to assess the integrity of elbows, are proposed. Those proposed solutions are expected to be the technical basis for revisions of CC N-513-4 through the validation.

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

• Journal of Software Assessment and Valuation
• /
• v.10 no.1
• /
• pp.19-26
• /
• 2014

In this paper, we propose a method for protecting Android applications against reverse engineering attacks. In this method, the server encrypts the original executable code (DEX) included in an APK file, inserts into the APK file a stub code that decrypts the encrypted DEX later at run-time, and distributes the modified APK file. The stub code includes an integrity validation code to detect attacks on itself. When a user installs and executes the APK file, the stub code verifies the integrity of itself, decrypts the encrypted DEX, and loads it dynamically to execute. Since the original DEX is distributed as an encrypted one, we can effectively protect the intellectual property. Further, by verifying the integrity of the stub code, we can prevent malicious users from bypassing our method. We applied the method to 15 Android apps, and evaluated its effectiveness. We confirmed that 13 out of them operates normally.

• Journal of the Korean Geotechnical Society
• /
• v.21 no.4
• /
• pp.51-63
• /
• 2005

Concrete pavement may suffer from material deterioration or structural problems, which lead to surface cracks and deflection of a concrete pavement. Degraded concrete pavement, when it is still under operation, should be recovered by an urgent maintenance to avoid the discontinued service leading to the significant traffic problems and economic loss. Seismic techniques are good tools to assess the structural integrity of concrete runway. It is because seismic techniques can evaluate engineering properties nondestructively and quickly and the evaluation can be extended to subgrade. In this study, a series of numerical simulations of stress-wave propagation were performed to verify feasibility of seismic techniques as an assessment tool. Based on the results of the numerical simulation, a framework of using seismic techniques was presented fur the nondestructive integrated assessment fur structural integrity of concrete runway. And the presented framework was applied to $\bigcirc\bigcirc$ concrete runway with surface cracks, which required urgent maintenance, to identify the causes of the surface cracks. The results obtained from the structural integrity assessment were compared with the measurements of the cores collected from the same runway for verification of the presented framework.