• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1213-1224
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• 2024

The liquid lead-bismuth cooled fast reactor has been in a single-phase, low-pressure, and high-temperature state for a long time during operation. Considering the requirement of calculation efficiency for long-term transient accident calculation, based on a homogeneous hydrodynamic model, one-dimensional heat conduction model, coolant flow and heat transfer model, neutron kinetics model, coolant and material properties model, this study used the fully implicit difference scheme algorithm of the convection-diffusion term to solve the basic conservation equation, to develop the transient analysis program NUSOL-LMR 2.0 for the lead-bismuth fast reactor system. The steady-state and typical design basis accidents (including reactivity introduction, loss of flow caused by main pump idling, excessive cooling, and plant power outage accidents) for the ABR have been analyzed. The results are compared with the international system analysis software ATHENA. The results indicate that the developed program can stably, accurately, and efficiently predict the transient accident response and safety characteristics of the lead-bismuth fast reactor system.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1687-1697
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• 2024

This study presents the Supporting platform for reactor fine flow characteristics calculation and analysis (Cilian platform), a user-friendly tool that supports the analysis and optimization of pressurized water reactor (PWR) cores with mixing vanes using computational fluid dynamics (CFD) computing. The Cilian platform allows for easy creation and optimization of PWR's main CFD calculation schemes and autonomously manages CFD calculation and analysis of PWR cores, reducing the need for human and computational resources. The platform's key features enable efficient simulation, rapid solution design, automatic calculation of core scheme options, and streamlined data extraction and processing techniques. The Cilian platform's capability to call external CFD software reduces the development time and cost while improving the accuracy and reliability of the results. In conclusion, the Cilian platform exemplifies an innovative solution for efficient computational fluid dynamics analysis of pressurized water reactor (PWR) cores. It holds great promise for driving advancements in nuclear power technology, enhancing the safety, efficiency, and cost-effectiveness of nuclear reactors. The platform adopts a modular design methodology, enabling the swift and accurate computation and analysis of diverse flow regions within core components. This design approach facilitates the seamless integration of multiple computational modules across various reactor types, providing a high degree of flexibility and reusability.

• Journal of Korean Society of Water and Wastewater
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• v.27 no.1
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• pp.21-29
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• 2013

Present study was conducted to evaluate the performance of Anaerobic Hybrid Reactor (AHR) combined with two types of anaerobic attached growth reactors at mesophilic temperature ($37^{\circ}C$). The reactor was operated at the influent substrate condition of 19,400 mg/L soluble chemical oxygen demand (sCOD). The organic loading rate (OLR) and flow rate were varied in the range of $9.5{\sim}22.5kg/m^3$. day and 10.6 ~ 26.0 L/day respectively since start-up was done. The COD removal efficiency of 93 % was measured at the OLR of $14kg/m^3$. day in AHR. However a reduction in removal efficiency to as low as 85 % could have been related to a combined effect of high concentration suspended solids (SS) concentration over 3,800 mg/L. On the other hand the total COD removal efficiencies were measured to be 96.3 % and 96.2 % for AHR+APF and AHR+ADF respectively. The pH of the POME was adjusted to neutral range by using sodium bicarbonate at the initial stages of the reactor feed, later stages pH adjustment was not required as the pH was maintained in the desired neutral range due to self-buffering capacity of the reactor. The reactor proved to be economically acceptable and operationally stable. The biogas was measured to have $CH_4$ and $CO_2$ with a ratio of 35:65, and methane gas production rate was estimated to be $0.17{\sim}10.269L\;CH_4/g\;COD_{removed}$.

• Preventive Nutrition and Food Science
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• v.4 no.1
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• pp.79-83
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• 1999

A potentiometric biosensor employing a CO3-2 ion-selective electrode(ISE) and malic enzyme immobilization in al flow injection analysis (FIA) system was constructed. Analytical parameters were optimized for L-malate determination . The CO3-2 -ISE-FIA system was composed of a pump, an injector, a malic enzyme (EC1.1.1.40) reactor, a CO3-2 ion-selective electrode, a pH/mV meter and a recorder. Cofactor NADP was also injected with substrate for theenzyme reaction into the system. Optimized analytical parameters for L-malate determination in the CO3-2 ISE-FIA system were as follows ; flow rate, 14.5ml/hr ; sample injection volume, 100ul; enzyme loading in the reactor, 20 units ; length of the enzyme reactor , 7 cm ; tubing length form the enzyme reactor to the detector as a geometric factor in FIA, 15 cm . The response time for measuring the entire L-malate concentration range (10-2 ~10-5 mol/L ; 4 injections )was <15minutes . In this CO3-2 -ISE-FIA system, the potential differences due to th eformation of CO3-2 by the reaction of malic enzyme on L-malate were correlated to L-malate concentration in the range of 10-2 ~10-5mol/L ; the detection limit was 10-5 mol/L. This potentionmetric CO3-2 ISE--FIA system was found to be useful for L-malate measurement.

• Proceedings of the Korean Society of Propulsion Engineers Conference
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• 2017.05a
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• pp.991-994
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• 2017

In hypersonic aircraft, increase of aerodynamic and engine heat lead thermal load in airframe. It could lead structural change of aircraft's component and malfunctioning. Endothermic fuels are liquid hydrocarbon fuels which are able to absorb the heat load by undergoing endothermic reactions. In this study, we investigated the method of measuring the heat sink of catalyst by using exo-tetrahydrodicyclopentadiene as a fuel in a packed bed flow reactor similar to the actual reaction conditions.

• Nuclear Engineering and Technology
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• v.41 no.6
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• pp.841-848
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• 2009

A simple model to analyze non-linear density-wave instability in a sodium-cooled helically coiled steam generator is developed. The model is formulated with three regions with moving boundaries. The homogeneous equilibrium flow model is used for the two-phase region and the shell-side energy conservation is also considered for the heat flux variation in each region. The proposed model is applied to the analysis of two-phase instability in a JAEA (Japan Atomic Energy Agency) 50MWt No.2 steam generator. The steady state results show that the proposed model accurately predicts the six cases of operating temperatures on the primary and secondary sides. The sizes of three regions, the secondary side pressure drop according to the flow rate, and the temperature variation in the vertical direction are also predicted well. The temporal variations of the inlet flow rate according to the throttling coefficient, the boiling and superheating boundaries and the pressure drop in the two-phase and superheating regions are obtained from the unsteady analysis.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.257-272
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• 2007

Supercritical water-cooled reactors (SCWRs) are recognized as a Generation IV reactor concept. The Super LWR is a pressure-vessel type thermal spectrum SCWR with downward-flow water rods and is currently under study at the University of Tokyo. This paper reviews Super LWR safety. The fundamental requirement for the Super LWR, which has a once-through coolant cycle, is the core coolant flow rate rather than the coolant inventory. Key safety characteristics of the Super LWR inhere in the design features and have been identified through a series of safety analyses. Although loss-of-flow is the most important abnormality, fuel rod heat-up is mitigated by the "heat sink" and "water source" effects of the water rods. Response of the reactor power against pressurization events is mild due to a small change in the average coolant density and flow stagnation of the once-through coolant cycle. These mild responses against transients and also reactivity feedbacks provide good inherent safety against anticipated-transient-without-scram (ATWS) events without alternative actions. Initiation of an automatic depressurization system provides effective heat removal from the fuel rods. An "in-vessel accumulator" effect of the reactor vessel top dome enhances the fuel rod cooling. This effect enlarges the safety margin for large LOCA.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3919-3927
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• 2022

This study presents an optimal design of the coolant system of a non-refueling full-life small reactor by analyzing the space-integrated geometrical and electromagnetic variables of an extra vessel electromagnetic pump (EVEMP) for the circulation of a lead-bismuth eutectic (LBE) coolant. The EVEMP is an ideal alternative to the thermal-hydraulic system of non-refueling full-life micro reactors as it possesses no internal structures, such as impellors or sealing structures, for the transportation of LBE. Typically, the LBE passes through the annular flow channel of a reactor, is cooled by the heat exchanger, and then circulates back to the EVEMP flow channel. This thermal-hydraulic flow method is similar to natural circulation, which enhances thermal efficiency, while providing a golden time for cooling cores in the event of an emergency. When the forced circulation technology of the EVEMP was applied, the non-refueling full-life micro reactor achieve an output power of 60 MWt, which is higher than that achievable via the natural circulation method (30 MWt). Accordingly, an optimized EVEMP for Micro URANUS with a flow rate of 4196 kg/s and developed pressure of 73 kPa under a working temperature of 250 ℃ was designed.

• Journal of computational fluids engineering
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• v.18 no.2
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• pp.1-8
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• 2013

In nuclear power plant, the reactor cooling system has maintained high-Reynolds-number flow above 1E+07 to cool a heat generated by the reactor. To minimize uncertainty for flow calibration, it is necessary to simulate the high Reynolds' number flow. Y-connection is selected to connect four (4) parallel high flow circulation pumps for maintaining the high flow rate. This paper describes the characteristics for Y-connection by computer flow simulation. It was confirmed through the results that the pressure loss of the Y-connection was lower than that of T-connection. Also as the connection angle of Y-connection was small, as the pressure loss was low.

• Journal of the Korean Society of Safety
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• v.20 no.2 s.70
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• pp.32-37
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• 2005

The purpose of this study is for the sensitivity study f9r a Steam Generator Tube Rupture (SGTR) of the System-integrated Modular Advanced ReacTor for a Pilot (SMART-P) plant. The thermal hydraulic analysis of a SGIR for the Limiting Conditions for Operation (LCO) is performed using TASS/SMR code. The TASS/SMR code can calculate the core power, pressure, flow, temperature and other values of the primary and secondary system for the various initiating conditions. The major concern of this sensitivity study is not the minimum Critical Heat Flux Ratio(CHFR) but the maximum leakage amount from the primary to secondary sides at the steam generator. Therefore the break area causing the maximum accumulated break flow is researched for this reason. In the case of a SGIR for the SMART-p, the total integrated break flow is 11,740kg in the worst case scenario, the minimum CHFR is maintained at Over 1.3 and the hottest fuel rod temperature is below 606"I during the transient. It means that the integrity of the fuel rod is guaranteed. The reactor coolant system and the secondary system pressures are maintained below 18.7MPa, which is system design pressure.