• Corrosion Science and Technology
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• v.15 no.2
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• pp.84-91
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• 2016

The transport and deposition of corrosion products in pressurized water nuclear reactor (PWR) steam generators have led to corrosion (SCC, denting etc.) problems. Lancing, mechanical cleaning and chemical cleaning have been used to reduce these problems. The methods of lancing and mechanical cleaning have limitations in removing corrosion products due to the structure of steam generator tubes. But high temperature chemical cleaning (HTCC) with EDTA is the most effective method to remove corrosion products regardless of the structure. However, EDTA in chemical cleaning aqueous solution and chemical cleaning environments affects the integrity of materials used in steam generators. The nuclear power plants have to perform the pre-test (also called as qualification test (QT)) that confirms the effect on the integrity of materials after HTCC. This is one of the series studies that assess the effect, and this study determines the effects of 20 % EDTA aqueous solution on defective tubes in high temperature chemical cleaning environments. The depth and magnitude of defects in steam generator (SG) tubes were measured by eddy current test (ECT) signals. Surface analysis and magnitude of defects were performed by using SEM/EDS. Corrosion rate was assessed by weight loss of specimens. The ECT signals (potential and depth %) of defective tubes increased marginally. But the lengths of defects, oxides on the surface and weights of specimens did not change. The average corrosion rate of standard corrosion specimens was negligible. But the surfaces on specimens showed traces of etching. The depth of etching showed a range on the nanometer. After comprehensive evaluation of all the results, it is concluded that 20 % EDTA aqueous solution in high temperature chemical cleaning environments does not have a negative effect on defective tubes.

• Nuclear Engineering and Technology
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• v.56 no.10
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• pp.4018-4030
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• 2024

The in-vessel retention through external reactor vessel cooling (IVR-ERVC) strategy is a key management strategy for early termination of a nuclear severe accident that can threaten the integrity of the reactor vessel. To simulate the physical phenomena of the molten corium, the smoothed particle hydrodynamic (SPH) method is utilized in this study. The SPH method is a Lagrangian computational fluid dynamic (CFD) method that can simulate multi-fluid stratification, turbulence, natural circulation, radiative heat transfer, thermal ablation, and crust formation. To address the external vessel cooling, it is coupled with a conventional 1-D nuclear system analysis method. The 1-D system analysis code can calculate the two-phase natural circulation of cooling water and the convective heat transfer on the external reactor vessel wall. These two simulation codes exchange the temperature and heat flux of the reactor vessel outer wall. This study numerically simulated the IVR-ERVC strategy for a Korean high-power reactor and compared it with the traditional lumped parameter method (LPM). Unlike LPM, this study provides localized detailed data about the thermal hydraulic behavior of molten corium and visualization of phenomena in the IVR-ERVC strategy. This enhances our understanding of the phenomena in IVR-ERVC strategy and introduces new perspectives.

• Journal of Korean Academy of Nursing
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• v.45 no.2
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• pp.240-250
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• 2015

Purpose: The purpose of this study was to validate the Korean version of the Ethical Leadership at Work questionnaire (K-ELW) that measures RNs' perceived ethical leadership of their nurse managers. Methods: The strong validation process suggested by Benson (1998), including translation and cultural adaptation stage, structural stage, and external stage, was used. Participants were 241 RNs who reported their perceived ethical leadership using both the pre-version of K-ELW and a previously known Ethical Leadership Scale, and interactional justice of their managers, as well as their own demographics, organizational commitment and organizational citizenship behavior. Data analyses included descriptive statistics, Pearson correlation coefficients, reliability coefficients, exploratory factor analysis, and confirmatory factor analysis. SPSS 19.0 and Amos 18.0 versions were used. Results: A modified K-ELW was developed from construct validity evidence and included 31 items in 7 domains: People orientation, task responsibility fairness, relationship fairness, power sharing, concern for sustainability, ethical guidance, and integrity. Convergent validity, discriminant validity, and concurrent validity were supported according to the correlation coefficients of the 7 domains with other measures. Conclusion: The results of this study provide preliminary evidence that the modified K-ELW can be adopted in Korean nursing organizations, and reliable and valid ethical leadership scores can be expected.

• Nuclear Engineering and Technology
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• v.33 no.3
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• pp.298-306
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• 2001

As fuel burnup proceeds, thermal gradients, differential swelling, and inter-assembly loading may induce assembly duct bowing. Since duct bowing affects the reactivity, such as long or short term power-reactivity-decrement variations, handling problem, caused by top end deflection of the bowed assembly duct, and the integrity of the assembly duct itself. Assembly duct bowing were first observed at EBR-ll in 1965, and then several designs of assembly ducts and core restraint system were used to accommodate this problem. In this study, NUBOW-2D KMOD was used to analyze the bowing behavior of the assembly duct under the KALIMER(Korea Advanced Liquid MEtal Reactor) core restraint system conditions. The mechanical behavior of assembly ducts related to several design parameters are evaluated. ACLP(Above Core Load Pad) positions, the gap distance between the ducts, and the gap distance between the duct and restraint ring were selected as the sensitivity parameter for the evaluation of duct deflection.

• Journal of the Korean Society of Safety
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• v.17 no.4
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• pp.52-60
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• 2002

Failure criterion is a parameter to represent the resistance to failure of locally wall thinned pipe, and it depends on material characteristics, defect geometry, applied loading type, and failure mode. Therefore, accurate prediction of integrity of wall thinned pipe requires a failure criterion adequately reflected the characteristics of defect shape and loading in the piping system. In the present study, the finite element analysis was performed and the results were compared with those of pipe experiment to develop a sound criterion for failure of internally wall thinned pipe subjected to combined pressure and bending loads. By comparing the predictions of failure to actual failure load and displacement, an appropriate criterion was investigated. From this investigation, it is concluded that true ultimate stress criterion is the most accurate to predict failure of wall thinned pipe under combined loads, but it is not conservative under some conditions. Engineering ultimate stress estimates the failure load and displacement reasonably for al conditions, although the predictions are less accurate compared with the results predicted by true ultimate stress criterion.

• Nuclear Engineering and Technology
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• v.53 no.9
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• pp.3112-3121
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• 2021

One of the major factors affecting the life span of a Reactor Pressure Vessel (RPV) is the Pressurised Thermal Shock (PTS). PTS is a thermo-mechanical load on the RPV wall due to steep temperature gradients and structural load created by internal pressure of the fluid within the RPV. Safe operating life of a nuclear power plant is ensured by carrying out fracture analysis of the RPV against thermal shock. Carrying out fracture tests on RPV/large scale components is not always feasible. Hence, studies on laboratory level specimens are necessary to validate and supplement the prototype results. This paper aims to study the fracture behaviour of standard Compact Tension [C(T)] specimens, made of RPV steel 20MnMoNi55, subjected to thermal shock through experimental and numerical investigations. Fracture tests have been carried out on the C(T) specimens subjected to thermal transient load and tensile load to quantify the effect of thermal shock. Crack resistance curves are obtained from the fracture tests as per ASTM E1820 and compared with those obtained numerically using XFEM and a good agreement was found. A quantitative study on the crack tip plastic zone, computed using cohesive segment approach, from the numerical analyses justified the experimental crack initiation toughness.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4030-4048
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• 2022

In this study, road transportation tests were conducted with surrogate fuel assemblies under normal conditions of transport to evaluate the vibration and shock load characteristics of spent nuclear fuel (SNF). The overall test data analysis was conducted based on the measured acceleration and strain data obtained from the speed bump, lane-change, deceleration, obstacle avoidance, and circular tests. Furthermore, representative shock response spectrums and power spectral densities of each test mode were acquired. Amplification or attenuation characteristics were investigated according to the load transfer path. The load attenuated significantly as it transferred from the trailer to the cask. By contrast, the load amplified as it transferred from the cask to the surrogate SNF assembly. The fuel loading location on the cask disk assembly did not exhibit a significant influence on the strain measured from the fuel rods. The principal strain was in the vertical direction, and relatively large strain values were obtained in spans with large spacing between spacer grids. The influence of the lateral location of fuel rods was also investigated. The fuel rods located at the side exhibited relatively large strain values than those located at the center. Based on the strain data obtained from the test results, a hypothetical road transportation scenario was established. A fatigue evaluation of the SNF rod was performed based on this scenario. The evaluation results indicate that no fatigue damage occurred on the fuel rods.

• KSCE Journal of Civil and Environmental Engineering Research
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• v.26 no.5A
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• pp.901-908
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• 2006

The structural integrity of containment building in Nuclear Power Plants has to be verified by the ISI(In Service Inspection) because there are some variations on the structural behavior of it due to the change of the physical properties of concrete and tendon with the lapse of time. In this study, the program 'SAPONC-CANDU' which can monitor and analyze the structural behavior of the containment building with grouted tendon (CANDU-type, 'Wolsong unit-2, 3, and 4' in Korea) was developed. This program is based on the algorithm which can calculate the prediction values of the quantities of strain variation for the vibrating-wire strain gauges embedded into the concrete of the containment building under temperature and time dependent factors which are creep, shrinkage, and prestressing force. The readings of the strain gauges are used as input data for the operation of the program. And it finally provides graphically a prediction value, line and band of the quantity of strain variation for the respective strain gauges, therefore, it is thought that the site engineers are able to assess the structural integrity of the domestic containment building with grouted tendon with easy using this program.

• 한국전산유체공학회:학술대회논문집
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• 2009.11a
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• pp.121-128
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• 2009

During a hypothetical high-pressure accident in a nuclear power plant (NPP), molten corium can be ejected through a breach of a reactor pressure vessel (RPV) and dispersed by a following jet of a high-pressure steam in the RPV. The dispersed corium is fragmented into smaller droplets in a reactor cavity of the NPP by the steam jet and released into other compartments of the NPP by a overpressure in the cavity. The fragments of the corium transfer thermal energy to the ambient air in the containment or interact chemically with steam and generate hydrogen which may be burnt in the containment. The thermal loads from the ejected molten corium on the containment which is called direct containment heating (DCH) can threaten the integrity of the containment. DCH in a NPP containment is related to many physical phenomena such as multi-phase hydrodynamics, thermodynamics and chemical process. In the evaluation of the DCH load, the melt dispersion rates depending on the RPV pressure are the most important parameter. Mostly, DCH was evaluated by using lumped-analysis codes with some correlations obtained from experiments for the dispersion rates. In this study, MC3D code was used to evaluate the dispersion rates in the APR1400 NPP during the high-pressure accidents. MC3D is a two-phase analysis code based on Eulerian four-fields for melt jet, melt droplets, gas and water. The dispersion rates of the corium melt depending on the RPV pressure were obtained from the MC3D analyses and the values specific to the APR1400 cavity geometry were compared to a currently available correlation.

• Nuclear Engineering and Technology
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• v.43 no.1
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• pp.63-74
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• 2011

The 600 MWe, pool-type, sodium-cooled, metallic fuel loaded KALIMER-600 (Korea Advanced LiquId MEtal Reactor, 600 MWe) has been conceptually designed with an emphasis on safety by self-regulating (inherent/intrinsic) negative reactivity feedback in the core. Its inherent safety under the ATWS (Anticipated Transient Without Scram) events was demonstrated in an earlier study. Initiating events of an HCDA (Hypothetical Core Disruptive Accident), however, also need to be analyzed for assessment of the margins in the current design. In this study, a hypothetical triple-fault accident, ULOF (Unprotected Loss Of Flow) with a reduced pump halving time, is investigated as an initiator of a core disruptive accident. A ULOF with insufficient primary pump inertia may cause core sodium boiling due to a power-to-flow mismatch. If the positive sodium reactivity resulting from this boiling is not compensated for by other intrinsic negative reactivity feedbacks, the resulting core power burst would challenge the fuel integrity. The present study focuses on determination of the limit of the pump inertia for assuring inherent reactivity feedback and behavior of the core after sodium boiling as well. Transient analyses are performed with the safety analysis code SSC-K, which now incorporates a new sodium boiling model. The results show that a halving time of more than 6.0 s does not allow sodium boiling even with very conservative assumptions. Boiling takes place for a halving time of 1.8 s, and its behavior can be predicted reasonably by the SSC-K.