• Nuclear Engineering and Technology
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• v.55 no.4
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• pp.1310-1323
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• 2023

A reactor coolant pump (RCP) is essential for transporting coolant in the primary loop of pressurized water reactors. In the advanced passive reactor, the absence of a long pipeline between the steam generator and RCP serves as a transition section, resulting in a non-uniform flow field at the pump inlet. Therefore, the characteristics of the pump should be investigated under non-uniform flow to determine its influence on the pump. In this study, the pressure pulsation characteristics were examined in the time and frequency domains, and the sources of low-frequency and high-amplitude signals were analyzed using wavelet coherence analysis and numerical simulation. From computational fluid dynamics (CFD) results, non-uniform inflow has a great effect on the flow structures in the pump's inlet. The pressure pulsation in the pump at the rated flow increased by 78-128.7% under the non-uniform inflow condition in comparison with that observed under the uniform inflow condition. Furthermore, a low-frequency signal with a high amplitude was observed, whose energy increased significantly under non-uniform flow. The wavelet coherence and CFD analysis verified that the source of this signal was the low-frequency pulsating vortex under the steam generator.

• Journal of Korean Society on Water Environment
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• v.25 no.6
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• pp.896-901
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• 2009

In this study, in order to apply the air and pure oxygen in the Jet Loop Reactor (JLB) in which the oxygen transfer rate is high, differentiate the operation mode according to each air flowrate and liquid flowrate and investigate the oxygen transfer characteristic, an experiment was carried out. The oxygen concentration with the air flowrate ($Q_g$) and liquid flowrate ($Q_L$) was identical but the oxygen transfer coefficient ($K_L{\cdot}a$) is linear depending on degree of two factors. The width of an increase is small in $0.1min^{-1}$ when the air flowrate is 0.2 L/min with increasing the liquid flowrate. Whereas, the increment was exposed to be very high for $1.5min^{-1}$ when the air flowrate was 5 L/min. In the experiments using the pure oxygen, it was 30 mg/L of oxygen concentration finally and it was 3.5 times than using the air. But the time reached the saturated concentration was similar to using the air, and $K_L{\cdot}a$ was similar to using the air too. Analysis between two independent variable and oxygen transfer of the correlation is the same model like $K_L{\cdot}a={0.0161Q_L}^{1.5371}{Q_g}^{0.5433}$ using with coefficient non linear regression analysis. It was resulted that the liquid flowrate were approximately three times than air flowrate on effect to oxygen transfer rate.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.33 no.2
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• pp.179-184
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• 2009

The Fuel Test Loop(FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR(Pressurized Water Reactor) and CANDU(CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI(Korea Atomic Energy Research Institute). It consists of In-Pile Section(IPS) and Out-of Pile System(OPS). The IPS, which is located inside the pool is divided into 3-parts; the in-pool pipes, the IVA(IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to keep the functionality of the reactor coolant pressure boundary. The IVA is manufactured by local company and the functional test and verification were done through pressure drop, vibration, hydraulic and leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by local technique and have finally tested under high temperature and high pressure. The IVA and piping did not experience leakage, as we have checked the piping, flanges, assembly parts. We have obtained good data during the three cycle test which includes a pressure test, pressure and temperature cycling, and constant temperature.

• Korean Chemical Engineering Research
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• v.58 no.4
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• pp.665-674
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• 2020

Low-frequency pressure fluctuations in an external-loop airlift reactor were investigated. Low-frequency pressure fluctuations could be measured by shooting videos about liquid levels in the four piezometric tubes which were installed at the lower and upper parts of the riser and downcomer using a cellular phone. The periodic characteristics of pressure fluctuations were proved by the calculation of their auto-correlation function and cross-correlation function. Even if the riser superficial gas velocity was constant, the riser and downcomer gas holdups as well as wall pressures were periodically changed due to the inertia of circulating liquid. In general, the intensity of pressure fluctuations increased with an increase in the gas velocity. When the unaerated liquid height was 0.04 m, the maximum period of pressure fluctuations was found at the specific gas velocity (0.14 ms^-1). It was because the maximum inertia of circulating liquid resulted from a reduction in the increasing rate of the liquid circulation velocity and a decrease in the volume of the effectively circulating liquid with an increase in the gas velocity.

• Journal of Energy Engineering
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• v.25 no.4
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• pp.85-92
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• 2016

Since the power density of the VHTR(Very High Temperature Reactor) is lower, there is less possibility of core melt. VHTR has no risk of explosion caused by hydrogen generation when the loss of coolant accident occurs, which is another advantage. Along with safety benefit, it can be used as a process heat supplier near demand facilities because coolant temperature is very high enough to be used for industrial purpose. In this paper, we designed the primary system using VHTR and the secondary system providing electricity and process heat. Based on that 350 MW thermal reactor proposed by NGNP(Next Generation Nuclear Part), we developed conceptual model that the IHX(Intermediate Heat Exchanger) loop transports 300 MW thermal energy to the secondary system. In addition, we analyzed thermodynamic behavior and performed the efficiency analysis and optimization study depending on major parameters.

• Proceedings of the KSME Conference
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• 2001.06a
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• pp.636-643
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• 2001

Interference between the crossover leg of the reactor coolant system (RCS) and the pipe whip restraints (PWR) has brought a degradation issue of the integrity of the Reactor Coolant System in Westinghouse type nuclear power plants (NPPs) of Korea. According to the gap Inspect ion carried out during planned overhaul (Year 2000), interference between the crossover leg and the PWR was found in each RCS loop. This plant has had the high vibration problem on the RC pump 'B'. The reason for the high vibration in the RC pump 'B' had been massively surveyed and it was found that the crossover leg of RCS contacted with the PWR in hot condition. Since the contact between the crossover leg and the PWR changes the dynamic characteristics of the piping system for the RCS, this is considered as one reason for the high vibration. And a possibility of overstress on the crossover leg due to the contact with the PWR should be evaluated. Through performing RCS integrity analyses, subsequent actions were initiated to increase the gap between those parts. As the results of the appropriate separation between two parts, it was reported that there was no unusual noise or vibration during plant heat-up. In this paper, the evaluations for the gap between the crossover leg and the PWR and the structural integrity due to loop binding is described.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.37 no.5
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• pp.665-671
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• 2013

In a Korean Generation IV prototype sodium-cooled fast reactor (SFR), various types of high-temperature heat exchangers such as IHX (intermediate heat exchanger), DHX (decay heat exchanger), AHX (air heat exchanger), FHX (finned-tube sodium-to-air heat exchanger), and SG (steam generator) are to be designed and installed. In this study, the high-temperature design and integrity evaluation of the sodium-to-air heat exchanger AHX in the STELLA-1 (sodium integral effect test loop for safety simulation and assessment) test loop already installed at KAERI (Korea Atomic Energy Research Institute) and FHX in the SEFLA (sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger) test loop to be installed at KAERI have been performed. Evaluations of creep-fatigue damage based on full 3D finite element analyses were conducted for the two heat exchangers according to the high-temperature design codes, and the integrity of the high-temperature design of the two heat exchangers was confirmed.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2789-2802
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• 2020

Small Modular Reactors (SMRs) are attracting wide attention due to their outstanding performance, extensive studies have been carried out for lead-based fast reactors (LFRs) that cooled with Lead or Lead-bismuth (LBE), and small modular natural circulation LFR is one of the promising candidates for SMRs and LFRs development. One of the challenges for the design small modular natural circulation LFR is to master the natural circulation thermal-hydraulic performance in the reactor primary circuit, while the natural circulation characteristics is a coupled thermal-hydraulic problem of the core thermal power, the primary loop layout and the operating state of secondary cooling system etc. Thus, accurate predicting the natural circulation LFRs thermal-hydraulic features are highly required for conducting reactor operating condition evaluate and Thermal hydraulic design optimization. In this study, a thermal-hydraulic analysis code is developed for small modular natural circulation LFRs, which is based on several mathematical models for natural circulation originally. A small modular natural circulation LBE cooled fast reactor named URANUS developed by Korea is chosen to assess the code's capability. Comparisons are performed to demonstrate the accuracy of the code by the calculation results of MARS, and the key thermal-hydraulic parameters agree fairly well with the MARS ones. As a typical application case, steady-state analyses were conducted to have an assessment of thermal-hydraulic behavior under nominal condition, and several parameters affecting natural circulation were evaluated. What's more, two characteristics parameters that used to analyze natural circulation LFRs natural circulation capacity were established. The analyses show that the core thermal power, thermal center difference and flow resistance is the main factors affecting the reactor natural circulation. Improving the core thermal power, increasing the thermal center difference and decreasing the flow resistance can significantly increase the reactor mass flow rate. Characteristics parameters can be used to quickly evaluate the natural circulation capacity of natural circulation LFR under normal operating conditions.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.9
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• pp.1241-1248
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• 2010

In large-scale production of hydrogen, a PHE (Process Heat Exchanger) is a key component because the heat required to carry out the Sulfur-Iodine chemical reaction that yields hydrogen is transferred from a VHTR (Very High Temperature Reactor) by the PHE. Korea Atomic Energy Research Institute established a helium gas loop for conducting performance test of components that are used in the VHTR. In this study, as a part of high-temperature structural-integrity evaluation of a designed PHE prototype that is scheduled to be tested in the helium gas loop, we carried out high-temperature structural-analysis modeling, thermal analysis, and thermal-expansion analysis for the designed PHE prototype. An appropriate constraint condition is proposed at the end of the in-flow and out-flow pipelines of the primary and secondary coolants and the proposed constraint condition will be applied to the design of the performance-test loop setup for the designed PHE prototype.

• Clean Technology
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• v.23 no.2
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• pp.205-212
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• 2017

We investigated the performance of a jet loop reactor (JLR) with the two-fluid venturi-type swirl nozzle (TVSN) during experiment for ammonia removal by air stripping from a synthetic wastewater, and compared it with that of a JLR with the two-fluid venturi-type conventional nozzle (TVCN), with the variation of pH, liquid circulation rate ($Q_L$), and air flow rate ($Q_G$). Their performance levels were compared based on the ammonia removal efficiency and overall mass transfer coefficient ($K_La$). Investigated parameters in a JLR were pH (10-12), air flow rate ($Q_G=5-20L\;min^{-1}$), and liquid circulation rate ($Q_L=25-35L\;min^{-1}$). Throughout the experiment, the ammonia removal efficiency and $K_La$ in a JLR with TVSN was higher than in a JLR with TVCN. This may be due to the enhanced turbulent intensity by swirling flow formed in the JLR with TVSN compared to that with TVCN. Further, we obtained higher $K_La$ when pH, $Q_L$ and $Q_G$ were increased. In particular, $K_La$ was increased more efficiently by increasing $Q_G$ than by increasing pH and $Q_L$.