• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
• /
• v.5 no.3
• /
• pp.199-212
• /
• 2007

If a type IP-2 transport package were to be subjected to a free drop test and a penetration test under the normal conditions of transport, it should prevent a loss or dispersal of the radioactive contents and a more than 20% increase in the maximum radiation level at any external surface of the package. In this paper, we suggested the analytic method to evaluate the structural safety of a type IP-2 transport package using a thick steel plate for a structure part and a bolt for tying a bolt. Using an analysis a loss or dispersal of the radioactive contents and a loss of shielding integrity were confirmed for two kinds of type IP-2 transport packages to transport radioactive waste drums from a waste facility to a temporary storage site in a nuclear power plant. Under the free drop condition the maximum average stress at the bolts and the maximum opening displacement of a lid were compared with the tensile stress of a bolt and the steps in a lid, which were made to avoid a streaming radiation in the shielding path, to evaluate a loss or dispersal of radioactive waste contents. Also a loss of shielding integrity was evaluated using the maximum decrease in a shielding thickness. To verify the impact dynamic analysis for free drop test condition and evaluate experimentally the safety of two kinds of type IP-2 transport packages, free drop tests were conducted with various drop directions. For the tests we examined the failure of bolts and the deformation of flange to evaluate a loss or dispersal of radioactive material and measured the shielding thickness using a ultrasonic thickness gauge to assess a loss of shielding integrity. The strains and accelerations acquired from tests were compared with those by analyses to verify the impact dynamic analysis. The analytic results were larger than the those of test so that the analysis showed the conservative results. Finally, we evaluated the safety of the type IP-2 transport package under the stacking test condition using a finite element analysis. Under the stacking test condition, the maximum Tresca stress of the shielding material was 1/3 of the yielding stress. Two kinds of a type IP-2 transport package were safe for the free drop test condition and the stacking test condition.

• Clinics in Shoulder and Elbow
• /
• v.12 no.2
• /
• pp.199-206
• /
• 2009

Purpose: Our goal for this study was to prospectively evaluate the functional & structural outcomes, by means of CT arthroscopy, of arthroscopic double-row fixation for treating rotator cuff tear. We also attempted to determine the variants that affect the functional & structural outcomes. Materials and Methods: Twenty seven consecutive patients underwent arthroscopic rotator cuff repair with double-row fixation. The average age at the time of the operation was fifty six years. The preoperative and postoperative examinations consisted of determining the Constant score, the score for the visual analogue scale for pain, the UCLA score, the American Shoulder and Elbow Surgeons (ASES) score, as well as a full physical examination of the shoulder. Preoperative MR arthrography was used to evaluate the integrity and atrophy of the rotator cuff. We measured the intraoperative tear size in the sagittal and coronal planes. Postoperative CT arthrography was used at one year postoperatively to evaluate the integrity and atrophy of the repaired tendons and muscles. Results: Preoperative MR arthrography revealed an average 29.22 mm tear size in the sagittal plane and an average 22.72 mm tear size in the coronal plane. Twelve cases of supraspinatus muscle atrophy and two cases of infraspinatus atrophy were observed on the preoperative MR arthrography. The average clinical outcome scores all significantly improved at the time of follow-up. At a mean of one year postoperatively, CT arthrography revealed 48.1% of the shoulders had healed, 11.1% showed incomplete healing and 40.7% showed retear of the repaired tendon. Conclusion: Arthroscopic double-row repair can result in improved clinical outcomes and good patient satisfaction. However, the problems about how to enhance healing of the repaired tendon still remain.

• Journal of the Korea Concrete Institute
• /
• v.24 no.3
• /
• pp.249-258
• /
• 2012

Integral and rigid frame bridges have advantages in bridge maintenance and structural efficiency by eliminating expansion joints and bridge supports. However, the detail of typical girder-abutment connection is rather complex and increases construction cost depending on construction detail. For the purpose of compensating disadvantages such as complexity and additional cost, a new type of bridge is proposed in this study, which improves the efficiency of construction by simplifying the construction detail of girder-abutment connection. The proposed bridge has the connection detail of steel girder and abutment integrated by prestressed PS bar installed in the connection. In this study, finite element analysis and fatigue load test are conducted to evaluate the fatigue capacity of the proposed girder-abutment connection. The results of the finite element analysis revealed that the possibility of the fatigue damage in the girder-abutment connection is very low. The results of the fatigue load test verified that the integrity of the girder and abutment connection is maintained after 2,000,000 cycles of fatigue loading.

• Journal of the Korean Recycled Construction Resources Institute
• /
• v.6 no.1
• /
• pp.66-71
• /
• 2018

In general, when an existing pier is enlarged and reinforced using a small diameter pile, bonded anchor with deformed reinforcing bars is used to maintain the integrity of the joint. However, in the case of bonded anchors, the performance depends largely on the type of joint material. Nevertheless, unlike mechanical anchors, there is no standard method for designing appropriate design methods and proper performance evaluation. Therefore, in this study, the performance of the anchoring anchor was evaluated by performing a model experiment using the reinforcing bars and anchor reinforcing bars. Experimental results show that the structural performance of the unbonded specimen is the best, and the failure mode is the punching shear failure. The deflection of the end of the member is smaller than that of the unconnected member, The deflection of the connected member is larger than the deflection of the small connected member. As the load increases, the possibility of slippage of the anchor steel or fold connection rebar is high.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.25 no.1
• /
• pp.61-69
• /
• 2001

EFP(Fitness For Purpose) type defect assessment methodologies based on ECA(Engineering Critical Analysis) have been established and are in use for the structural integrity evaluation of gas pipelines. ECA usually includes the fracture mechanics analysis, and it is assumed that the J-integral uniquely characterizes the crack-tip stress-strain field. However, it has been proven that the J-integral alone can not be sufficient to characterize the crack-tip field under low levels of constraint with a single parameter. Since pipeline structures are made of ductile material, locally loaded in tension, cracks may experience low level of constraint, and therefore, J-dominance will be lost. For this reason, the level of constraint must be quantified to establish a precise assessment procedure for pipeline defects. The objective of this paper is to investigate the fracture behavior of a crack in gas pipeline(KS D 3507) by quantifying the level of constraint. For this purpose, tensile tests and CTOD tests were performed at room temperature(24$\^{C}$) and low temperature(-40$\^{C}$) to obtain the material properties. J-Q analyses were performed for SENB and SENT specimens based on 2-D finite element analyses, in order to investigate the in-plane constraint effects on pipeline defects. For precise assessment of cracks, especially shallow cracks, in KS D 3507 pipeline, constraint effect must be considered.

• Transactions of the Korean Society of Mechanical Engineers A
• /
• v.39 no.5
• /
• pp.453-460
• /
• 2015

Since the Fukushima nuclear accident, several researchers are extensively studying the effect of torsion on the piping systems In nuclear power plants. Piping installations in power plants with a circumferential crack can be operated under combined loading conditions such as bending and torsion. ASME Code provides flaw evaluations for fully plastic fractures using limit load criteria for the structural integrity of the cracked pipes. According to the recent version of Code, combined loadings are provided only for the membrane and bending. Even though actual operating conditions have torsion loading, the methodology for evaluating torsion load is not established. This paper provides the results of limit load analyses by using finite element models for circumferentially cracked pipes under pure bending, pure torsion, and combined bending and torsion with tension. Theoretical limit load solutions based on net-section fully plastic criteria are suggested and verified with the results of finite element analyses.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.21 no.2
• /
• pp.197-203
• /
• 2001

Micro-defects induced by design and production failure or working environments are known as the cause of SCC(Stress Corrosion Cracking) in aged structures. Therefore, the evaluation of structural integrity based on micro-cracks is required not only a manufacturing step but also in-service term. So we introduce a new nondestructive inspection method using the magneto-optical film to detect micro-cracks. The method has some advantage such as high testing speed, real time data acquistion and the possibility of remote sensing by using of a magneto-optical film that takes advantage of the change of magnetic domains and domain walls. This paper introduces the concept of the new nondestructive inspection method using the magneto-optical film, also proves the possibility of this method as a remote testing system under oscillating load considering application on real fields by applying the method to four types of specimens.

• Journal of the Korean Society for Nondestructive Testing
• /
• v.31 no.5
• /
• pp.543-551
• /
• 2011

Steam generator tubes for nuclear power plants contain the local shape transitions on their inner or outer surface such as dent, bulge, over-expansion, eccentricity, deflection, and so on by the application of physical force during the tube manufacturing and steam generator assembling and by the sludge (that is, corrosion products) produced during the plant operation. The structural integrity of tubes will be degraded by generating the corrosive crack at that location. The profilometry using the traditional bobbin probes which are currently applied for measuring the profile change of tubes gives us basic information such as axial locations and average magnitudes of deformations. However, the three-dimensional quantitative evaluation on circumferential locations, distributional angle, and size of deformations will have to be conducted to understand the effects of residual stresses increased by local deformations on corrosive cracking of tubes. Steam generator tubes of Korean standard nuclear power plants expanded within their tube-sheets by the explosive expansion method and suffered from corrosive cracks in the early stage of power operation. Thus, local deformations of steam generator tubes at the top of tube-sheet were measured with an advanced rotating probe and a laser profiling system for the two cases where the tubes expanded by the explosive expansion method and hydraulic expansion. Also, the trends of eccentricity, deflection, and over-expansion of tubes were evaluated. The advanced eddy current profilometry was confirmed to provide accurate information of local deformations compared with laser profilometry.

• Transactions of the Korean Society of Pressure Vessels and Piping
• /
• v.12 no.1
• /
• pp.62-69
• /
• 2016

Current guidance considers that uniaxially loaded specimen with a deep crack is used for the determination of the ductile-to-brittle transition temperature. However, reactor pressure vessel is under biaxial loading in real and the existence of deep crack is not probable through periodic in-service-inspection. The elastic stress intensity factor and the elastic-plastic J-integral which were used for crack-tip stress field and fracture mechanics assessment parameters. The difference of the loading condition and crack geometry can significantly influence on these parameters. Thus, a constraint effect caused by differences between standard specimens and a real structure can over/underestimate the fracture toughness, and it affects the results of the structural integrity assessment, consequentially. The present paper investigates the constraint effects by evaluating the master curve $T_0$ reference temperature of PCVN (Pre-cracked Charpy V-Notch) and small scale cruciform specimens which was designed to simulate biaxial loading condition with shallow crack through the fracture toughness tests and 3-dimensional elastic-plastic finite element analyses. Based on the finite element analysis results, the fracture toughness values of a small scale cruciform specimen were estimated, and the geometry-dependent factors of the cruciform specimen considered in the present study were determined. Finally, the transferability of the test results of these specimens was discussed.

• Nuclear Engineering and Technology
• /
• v.24 no.1
• /
• pp.52-62
• /
• 1992

Radioactive nuclides deposited on out-of-core surface after the radiation in the core by the transport of corrosion products (CRUD) through the primary coolant system in PWR which is the major plant type in Korea, are leading sources of radiation exposure to plant maintenance personnel. Thus, the optimal chemistry operation method is required for the reduction of radiation exposure by the corrosion products. This study analysed the actual water chemistry operation data of four operating domestic PWRs. And in order to evaluate the coolant chemistry operation data, a computer code which can calculate the activity buildup in the various chemistry conditions of PWR coolant was employed. Through the analysis of comparison between the activity buildup of actual water chemistry operation mode and that of assumed Elevated Li operation mode calculated by the computer code, it was found that the out-of core radioactivity can be reduced by diminishing the deposition of corrosion products on the core in case that the Elevated Li operation mode is applied to the coolant chemistry operation of PWR. And the higher coolant pH operation was shown to have the advantage of the reduction of out-of-core activity buildup if the integrity of system structural materials and fuel cladding is guaranteed.