• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.59-68
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• 2020

ECCMIX component was introduced in RELAP5/MOD3 for calculating the interfacial condensation. Compared to other existing components in RELAP5, user experience of ECCMIX component is restricted to developmental assessment applications. To evaluate the capability of the ECCMIX component, ECCS experiment was conducted which included single-phase and two-phase thermal mixing. The experiment was carried out with test sections containing a main pipe (70 mm inner diameter) and a branch pipe (21 mm inner diameter) under the atmospheric pressure. The steam mass flow in the main pipe ranged from 0 to 0.0347 kg/s, and the subcooled water mass flow in the branch pipe ranged from 0.0278 to 0.1389 kg/s. The comparison of the experimental data with the calculation results illuminated that although the ECCMIX component was more difficult to converge than Branch component, it was a more appropriate manner to simulate interfacial condensation under two-phase thermal mixing circumstance, while the two components had no differences under single-phase circumstance.

• Nuclear Engineering and Technology
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• v.22 no.1
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• pp.12-19
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• 1990

The Loss of Offsite Power Transient at 77.5% power which occurred on June 9, 1981 at the Kori Unit 1 PWR (Pressurized Water Reactor) is simulated using the RELAP5/MOD2 system thermal-hydraulics computer code. Major thermal-hydraulic parameters are compared with the available plant data. The comparison of the analysis results with the plant data demonstrates that the RELAP5/MOD2 code has the capability to simulate the thermal-hydraulic behaviour of PWRs under accident conditions of this type with accuracy, except the pressurizer pressure and level. The pressurizer pressure increase is sensitive to the in surge now it is believed that the interracial heat transfer in a horizontal stratified flow regime may be estimated low and the compression effect due to insurge flow may be high. In the nodalization sensitivity study it is found that S/G noding with junctions between bypass plenum and steam dome is preferred to simulate the S/G water level decreasing and avoid the spurious level peak at trubine trip.

• Journal of Conservation Science
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• v.25 no.4
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• pp.373-381
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• 2009

To preserve steel cultural properties on display outdoors, the surface of these relics is processed with a covering treatment, which is a typical anti-corrosion method. Since the registered cultural heritage of Korea No. 78 Locomotive at Jangdan Station of Gyeongui Line will be treated by preservation process and displayed in the open outdoor area, effective surface covering agents were selected through testing to prevent the corrosion of the locomotive. The mixed wax are based on two types of microcrystalline wax which is one of regular surface covering agents for locomotives and bee‘s wax. For the weather resistance test, 3 types of wax were applied to each specimen of the locomotive and the specimens were tested in the Xenon-Arc Accelerated Weathering Tester for 1000 hours in accordance with the ISO 4892-1981 criteria and the test results were compared. As a result, on the surface of the specimen coated with the bee's wax, a whitening event appears after 400 hours which causes aging, and the specimens coated by the microcrystalline wax showed signs of a whitening event after 800 hours. The moisture contact angle with the microcrystalline wax coating was also less than with the bee's wax. Therefore, it was found that the high temperature microcrystalline wax is the most effective coating agent in terms of weather resistance and moisture blocking capability and the high temperature microcrystalline wax was adopted for coating the surface of the locomotive.

• Applied Chemistry for Engineering
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• v.10 no.4
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• pp.543-547
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• 1999

The utilization of fly-ash containing high levels of the unburned carbon was investigated. In this study adsorbents were manufactured from fly-ash ad a raw material and the manufactured adsorbents were applied to the waste-water treatment including heavy metals. Varying the anthracite(Jangsung coal) content, three types of pellet were made. The carbon content of pellet increased appreciably upon the addition of anthracite. After carbonization and activation using the pellets, adsorbents showed following characteristics; the range of hardness was between 85% and 96%, iodine number was from 100 mg/g to 300 mg/g. In proportion to the anthracite addition, hardness and iodine number increased. Through the adsorption experiments of heavy metals, removal efficiencies of Pb and Cr by manufactured adsorbents were over 90%. In case of fly-ash, removal efficiencies of Pb and Cr were 31.5% and 5.6% respectively. The reason why removal capacities of manufactured adsorbents were higher than fly-ash was postulated that adsorption capability was improved by unburned carbon and $SiO_2$ which included in fly-ash during steam activation.

• Nuclear Engineering and Technology
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• v.50 no.3
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• pp.356-367
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• 2018

A series of tests dedicated to station blackout (SBO) accident scenarios have been recently performed at the $Prim{\ddot{a}}rkreislauf-Versuchsanlage$ (primary coolant loop test facility; PKL) facility in the framework of the OECD/NEA PKL-3 project. These investigations address current safety issues related to beyond design basis accident transients with significant core heat up. This work presents a detailed analysis using the best estimate thermal-hydraulic code TRACE (v5.0 Patch4) of different SBO scenarios conducted at the PKL facility; failures of high- and low-pressure safety injection systems together with steam generator (SG) feedwater supply are considered, thus calling for adequate accident management actions and timely implementation of alternative emergency cooling procedures to prevent core meltdown. The presented analysis evaluates the capability of the applied TRACE model of the PKL facility to correctly capture the sequences of events in the different SBO scenarios, namely the SBO tests H2.1, H2.2 run 1 and H2.2 run 2, including symmetric or asymmetric secondary side depressurization, primary side depressurization, accumulator (ACC) injection in the cold legs and secondary side feeding with mobile pump and/or primary side emergency core coolant injection from the fuel pool cooling pump. This study is focused specifically on the prediction of the core exit temperature, which drives the execution of the most relevant accident management actions. This work presents, in particular, the key improvements made to the TRACE model that helped to improve the code predictions, including the modeling of dynamical heat losses, the nodalization of SGs' heat exchanger tubes and the ACCs. Another relevant aspect of this work is to evaluate how well the model simulations of the three different scenarios qualitatively and quantitatively capture the trends and results exhibited by the actual experiments. For instance, how the number of SGs considered for secondary side depressurization affects the heat transfer from primary side; how the discharge capacity of the pressurizer relief valve affects the dynamics of the transient; how ACC initial pressure and nitrogen release affect the grace time between ACC injection and subsequent core heat up; and how well the alternative feeding modes of the secondary and/or primary side with mobile injection pumps affect core quenching and ensure stable long-term core cooling under controlled boiling conditions.

• Journal of Korea Port Economic Association
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• v.25 no.3
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• pp.67-92
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• 2009

A tugboat (tug) is a boat that maneuvers vessels by pushing or towing them. Tugs move vessels that should not move themselves alone, such as ships in a crowded harbor or a narrow canal, or those that cannot move themselves, such as barges, disabled ships, or oil platforms. Tugboats are powerful for their size and strongly built, some are ocean-going. Historically tugboats were the first seagoing vessels to receive steam propulsion, freedom from the restraint of the wind, and capability of going in any direction. As such, they were employed in harbors to assist ships in docking and departure. Towage is in essence a service by one vessel to another vessel for a fixed remuneration. The most common reason for requiring this service is the lack of its own motive power. Conventionally, towage is defined as "the employment of one vessel to expedite the voyage of another, when nothing more is required than the accelerating of her progress". Apart from accelerating vessels, acquiring towage service is a common practice for towing barges, platform of drilling oil, floating ship yards, etc.

• Nuclear Engineering and Technology
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• v.27 no.6
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• pp.859-869
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• 1995

The present study is to understand the physical phenomena anticipated during the accident with RHR loss under mid-loop operation in a PWR and, at the same time, to examine the prediction capability of RELAP5/MOD3.1 on such an accident, by simulating an integral test relevant to this accident for reliable analysis in an actual PWR. The selected experiment, i.g. BETHSY Test 6.9a, represents the configuration with the pressurizer manway open and steam generators unavailable during the accident. Accordingly, the results of this ok are sure to contribute to understanding both the key events as well as the sensitive parameters, anticipated in the accident, for validity of the actual analysis. In the simulation result overall behavior as well as major phenomena observed in the experiment have been predicted reasonably by RELAP5/MOD3.1, however, the problem associated with enormous computing time .due to small time step size has been encountered. Besides, the code prediction of higher swollen level in the pressure vessel has given rise to overestimation of both pressurizer level and RCS pressure. Subsequently, overprediction of the break flow through the manway has led to earlier core uncovery than that in the experiment by about 400 seconds. As a whole, it is demonstrated from both the experiment and the analysis that gravity feed has not been sufficient to recover the core level and thus additional forced feed has been necessary in this configuration.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.803-811
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• 2024

Nuclear accidents such as Fukushima Daiichi have highlighted the potential of passive safety systems to replace or complement active safety systems as part of the overall prevention and/or mitigation strategies. In addition, passive systems are key features of Small Modular Reactors (SMRs), for which they are becoming almost unavoidable and are part of the basic design of many reactors available in today's nuclear market. Nevertheless, their potential to significantly increase the safety of nuclear power plants still needs to be strengthened, in particular the ability of computer codes to determine their performance and reliability in industrial applications and support the safety demonstration. The PASTELS project (September 2020-February 2024), funded by the European Commission "Euratom H2020" programme, is devoted to the study of passive systems relying on natural circulation. The project focuses on two types, namely the SAfety COndenser (SACO) for the evacuation of the core residual power and the Containment Wall Condenser (CWC) for the reduction of heat and pressure in the containment vessel in case of accident. A specific design for each of these systems is being investigated in the project. Firstly, a straight vertical pool type of SACO has been implemented on the Framatome's PKL loop at Erlangen. It represents a tube bundle type heat exchanger that transfers heat from the secondary circuit to the water pool in which it is immersed by condensing the vapour generated in the steam generator. Secondly, the project relies on the CWC installed on the PASI test loop at LUT University in Finland. This facility reproduces the thermal-hydraulic behaviour of a Passive Containment Cooling System (PCCS) mainly composed of a CWC, a heat exchanger in the containment vessel connected to a water tank at atmospheric pressure outside the vessel which represents the ultimate heat sink. Several activities are carried out within the framework of the project. Different tests are conducted on these integral test facilities to produce new and relevant experimental data allowing to better characterize the physical behaviours and the performances of these systems for various thermo-hydraulic conditions. These test programmes are simulated by different codes acting at different scales, mainly system and CFD codes. New "system/CFD" coupling approaches are also considered to evaluate their potential to benefit both from the accuracy of CFD in regions where local 3D effects are dominant and system codes whose computational speed, robustness and general level of physical validation are particularly appreciated in industrial studies. In parallel, the project includes the study of single and two-phase natural circulation loops through a bibliographical study and the simulations of the PERSEO and HERO-2 experimental facilities. After a synthetic presentation of the project and its objectives, this article provides the reader with findings related to the physical analysis of the test results obtained on the PKL and PASI installations as well an overall evaluation of the capability of the different numerical tools to simulate passive systems.