• Nuclear Engineering and Technology
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• v.50 no.6
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• pp.829-841
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• 2018

An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LSTF), which simulated a 17% hot leg intermediate-break loss-of-coolant accident in a pressurized water reactor (PWR). In the LSTF test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on the upper core plate in the upper plenum. Results of the uncertainty analysis with RELAP5/MOD3.3 code clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges. For studying the scaling problems to extrapolate thermal-hydraulic phenomena observed in scaled-down facilities, an experiment was performed for the OECD/NEA PKL-3 Project with the Primarkreislaufe Versuchsanlage (PKL), as a counterpart to a previous LSTF test. The LSTF test simulated a PWR 1% hot leg small-break loss-of-coolant accident with steam generator secondary-side depressurization as an accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for the primary pressure, the core collapsed liquid level, and the cladding surface temperature probably due to effects of differences between the LSTF and the PKL in configuration, geometry, and volumetric size.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3166-3175
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• 2022

The most numerical investigations of the thermal-hydraulic phenomena following the loss of the residual heat removal capability during the mid-loop operation of the pressurized water reactor were performed according to simplifications and are not sufficiently accurate. To perform more accurate and more reliable predictions of thermal-hydraulic accidents in a nuclear power plant using computational fluid dynamics codes, a more detailed methodology is needed. Modelling results identified that thermal stratification and natural convection are observed. Temperatures of lower monitoring points remain low, while temperatures of upper monitoring points increase over time. The water in the heated region, in the upper unheated region and the pipe region was well mixed due to natural convection, meanwhile, there is no natural convection in the lower unheated region. Water temperature in the pipe region increased after a certain time delay due to circulation of flow induced by natural convection in the heated and upper unheated regions. The modelling results correspond to the experimental data. The developed computational fluid dynamics methodology could be applied for modelling of two-component single/two-phase natural convection and thermal stratification phenomena during the mid-loop operation of the pressurized water reactor or other nuclear and non-nuclear installations at similar conditions.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1732-1741
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• 2020

The structural integrity of welded joints in the reactor pressure vessel (RPV) is directly related to the safety of nuclear power plants. The RPV is made from SA508-III steel in a pressurized water reactor. In this study, we investigated the effects of temperature on the tensile and fracture toughness properties of Chinese SA508-III welded joint in different sampling areas in order to provide reference data for structural integrity assessments of RPVs. The specimens used in tensile and fracture toughness tests were fabricated from the base metal (BM), weld metal (WM), and the heat-affected zone (HAZ) in the welded joint. The representative testing temperatures included the ambient temperature (20 ℃), upper shelf temperature (100 ℃), and service temperature (320 ℃). The results showed that temperature greatly affected the fracture toughness (JIC) values for the SA508-III welded joint. The JIC values for BM and HAZ both decreased remarkably from 20 ℃ to 320 ℃. The fracture morphologies showed that the BM and HAZ in the welded joint exhibited fully ductile fracture at 20 ℃, whereas partial cleavage fracture was mixed in ductile fracture mode at 100 ℃ and 320 ℃. The WM exhibited the ductile and cleavage fracture mixed mode at various temperatures, and the JIC values showed slight changes.

• Nuclear Engineering and Technology
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• v.49 no.4
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• pp.663-667
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• 2017

Delayed hydride cracking (DHC) was first observed in pressure tubes in Canadian CANDU reactors. In light water reactors, DHC was not observed until the late 1990s in high-burnup boiling water reactor (BWR) fuel cladding. In recent years, the focus on DHC has resurfaced in light of the increased interest in the cladding integrity during interim conditions. In principle, all spent fuel in the wet pools has sufficient hydrogen content for DHC to operate below $300^{\circ}C$. It is therefore of importance to establish the critical parameters for DHC to operate. This work studies the threshold stress intensity factor ($K_{IH}$) to initiate DHC as a function of temperature in Zry-4 for temperatures between $227^{\circ}C$ and $315^{\circ}C$. The experimental technique used in this study was the pin-loading testing technique. To determine the $K_{IH}$, an unloading method was used where the load was successively reduced in a stepwise manner until no cracking was observed during 24 hours. The results showed that there was moderate temperature behavior at lower temperatures. Around $300^{\circ}C$, there was a sharp increase in $K_{IH}$ indicating the upper temperature limit for DHC. The value for $K_{IH}$ at $227^{\circ}C$ was determined to be $2.6{\pm}0.3MPa$ ${\surd}$m.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.24 no.8 s.179
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• pp.2083-2090
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• 2000

In order to manage nuclear power plants safely and cost effectively, it is necessary to develop integrity evaluation methodologies for the main components. Recently, the integrity evaluation techniques were broadly studied regarding the license renewal of nuclear power plants which were approaching their design lives. Since the integrity evaluation process requires special knowledges and complicated calculation procedures, it has been allowed only to experts in the specified area. In this paper, an integrity evaluation system for reactor pressure vessel was developed. RVIES(Reactor Vessel Integrity Evaluation System) provides four specific integrity evaluation procedures covering PTS(Pressurized Thermal Shock) analysis, P-T(Pressure-Temperature) limit curve generation, USE(Upper Shelf Energy) analysis and Fatigue analysis. Each module was verified by comparing with published results.

• Korean Journal of Metals and Materials
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• v.50 no.10
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• pp.745-751
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• 2012

The Kroll process for magnesium reduction of titanium tetrachloride is used for mass production of titanium sponge. The present study was conducted in a lab scale reactor to develop a better understanding of the mechanism of titanium sponge formation in the Kroll reactor with respect to reaction degrees and the feeding rate of $TiCl_4$. The $MgCl_2$ produced during the initial stage of the reaction was not sunk into the molten magnesium, but covered the surface of the molten magnesium. As a result, subsequently fed $TiCl_4$ reacted with Mg exposed on the edge of molten $MgCl_2$ in the crucible. Therefore, titanium sponge grew toward the center of the crucible from the edge. The temperature of the molten magnesium increased remarkably with the increasing feeding rate of $TiCl_4$. Consequently, fed $TiCl_4$ reacted at the upper side of the crucible with evaporated Mg, and produced titanium on the upper surface of the crucible wall, which increased considerably with the feeding rate of $TiCl_4$.

• Journal of environmental and Sanitary engineering
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• v.12 no.3
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• pp.9-17
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• 1997

This research aims to remove nitrogen in the piggery wastewater by combined process with upflow anaerobic sludge blanket (UASB) and biofilm process. For the effective denitrification. anaerobic and anoxic reactors were connected to a reactor. The effluent of aerobix reactor was recycled equally with influent in the upper filter of anaerobic reactor for denitrification and outlet of UBF reactor was connected to the settling tank with $1.5{\;}{\ell}$ capacity and the settling sludge was repeatedly recycled to UASB zone. The organic loading rate of total reactor was operated from 0.4 to $3.1kgCOD/m^{3}/d$ and it was observed that the removal rate of TCOD was 80 to 95 percentage. Ammonia nitrogen was removed over 90 percentage in the less volumetric loading rate than $0.1{\;}kgN/m^{3}/d$. But because of non-limitation of organic materials, it was reduced to 70 percentage in the more volumetric loading rate than $0.6{\;}kgN/m^{3}/d$. But denitrification rate was observed 100 percentage in the all of loading rate. This is caused by the maintenance of optimum temperature, sufficient carbon source, and competition of electron acceptors. The results of COD mass balance at the $1.21{\;}kgCOD/m^{3}/d$ was observed with the 71.7% percentage of influent COD. It was revealed that the most part of organic materials was removed in the aerobic and the anaerobic reactor because 38.4 percentage was conversed into $CH_{4}$ gas and 11 percentage was removed in the aerobic reactor with cell synthesis and metabolism. Besides, 5.7% organics was used to denitrification reaction and 3.7% organics related to sulfate reduction.

• 한국전산유체공학회:학술대회논문집
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• 2005.10a
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• pp.105-110
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• 2005

Evaluation of elliptic blending turbulence model (EBM) together with the two-layer model, shear stress transport (SST) model and elliptic relaxation model (V2-F) is performed for a better prediction of thermal stratification in an upper plenum of a liquid metal reactor by applying them to the experiment conducted at JNC. The algebraic flux model is used for treating the turbulent heat flux. There exist much differences between turbulence models in predicting the temporal variation of temperature. The V2-F model and the EBM better predict the steep gradient of temperature at the interface of thermal stratification, and the V2-F model and EBM predict properly the oscillation of temperature. The two-layer model and SST model fail to predict the temporal oscillation of temperature.

• 한국전산유체공학회:학술대회논문집
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• 2004.10a
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• pp.221-225
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• 2004

Evaluation of turbulence models is performed for a better prediction of thermal stratification in an upper plenum of a liquid metal reactor by applying them to the experiment conducted at JNC. The turbulence models tested in the present study are the two-layer model, the $\kappa-\omega$ model, the v2-f model and the low-Reynolds number differential stress-flux model. When the algebraic flux model or differential flux model are used for treating the turbulent heat flux, there exist little differences between turbulence models in predicting the temporal variation of temperature. However, the v2-f model and the low-Reynolds number differential stress-flux model better predict the steep gradient o( temperature at the interface of thermal stratification, and only the v2-f model predicts properly the oscillation of temperature. The LES Is needed for a better prediction of the amplitude and frequency of the temperature fluctuation.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.6
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• pp.637-647
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• 2014

In pressurized water reactors, the upper head of the reactor pressure vessel (RPV) contains numerous control rod drive mechanism (CRDM) nozzles. These nozzles are fabricated by welding after being inserted into the RPV head with a room temperature shrink fit. The tensile residual stresses caused by this welding are a major factor in primary water stress corrosion cracking (PWSCC). Over the last 15 years, the incidences of cracking in alloy 600 CRDM nozzles have increased significantly. These cracks are caused by PWSCC and have been shown to be driven by the welding residual stresses and operational stresses in the weld region. Various measures are being sought to overcome these problems. The defects resulting from the welding process are often the cause of PWSCC acceleration. Therefore, any weld defects found in the RPV manufacturing process are immediately repaired by repair welding. Detailed finite-element simulations for the Korea Nuclear Reactor Pressure Vessel were conducted in order to predict the magnitudes of the repair weld residual stresses in the tube materials.