• Nuclear Engineering and Technology
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• v.42 no.6
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• pp.656-661
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• 2010

Lead-alloys are very attractive nuclear coolants due to their thermo-hydraulic, chemical, and neutronic properties. By utilizing the HELIOS (Heavy Eutectic liquid metal Loop for Integral test of Operability and Safety of PEACER$^2$) facility, a thermal hydraulic benchmarking study has been conducted for the prediction of pressure loss in lead-alloy cooled advanced nuclear energy systems (LACANES). The loop has several complex components that cannot be readily characterized with available pressure loss correlations. Among these components is the core, composed of a vessel, a barrel, heaters separated by complex spacers, and the plenum. Due to the complex shape of the core, its pressure loss is comparable to that of the rest of the loop. Detailed CFD simulations employing different CFD codes are used to determine the pressure loss, and it is found that the spacers contribute to nearly 90 percent of the total pressure loss. In the system codes, spacers are usually accounted for; however, due to the lack of correlations for the exact spacer geometry, the accuracy of models relies strongly on assumptions used for modeling spacers. CFD can be used to determine an appropriate correlation. However, application of CFD also requires careful choice of turbulence models and numerical meshes, which are selected based on extensive experience with liquid metal flow simulations for the KALLA lab. In this paper consistent results of CFX and Star-CD are obtained and compared to measured data. Measured data of the pressure loss of the core are obtained with a differential pressure transducer located between the core inlet and outlet at a flow rate of 13.57kg/s.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2699-2708
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• 2020

Helical-coil steam generator (HCSG) technology is a major design candidate for small modular reactors due to its compactness and capability to produce superheated steam with high generation efficiency. In this paper, we investigate the feasibility of the passively autonomous power maneuvering by coupling the 3-D transient multi-physics of a soluble-boron-free (SBF) core with a time-dependent HCSG model. The predictor corrector quasi-static method was used to reduce the cost of the transient 3-D neutronic solution. In the numerical system simulations, the feedwater flow rate to the secondary of the HCSGs is adjusted to extract the demanded power from the primary loop. This varies the coolant temperature at the inlet of the SBF core, which governs the passively autonomous power maneuvering due to the strongly negative coolant reactivity feedback. Here, we simulate a 100-50-100 load-follow operation with a 5%/minute power ramping speed to investigate the feasibility of the passively autonomous load-follow in a 450 MW_th SBF PWR. In addition, the passively autonomous frequency control operation is investigated. The various system models are coupled, and they are solved by an in-house Fortran-95 code. The results of this work demonstrate constant steam temperature in the secondary side and limited variation of the primary coolant temperature. Meanwhile, the variations of the core axial shape index and the core power peaking are sufficiently small.

• Nuclear Engineering and Technology
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• v.32 no.1
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• pp.57-66
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• 2000

The RELAP5 code has been developed for best-estimate simulation of transients and accidents for pressurized water reactors and their associated systems, but it has not been fully assessed for those of CANDU reactors. However, a previous study suggested that the RELAP5 code could be applicable to simulate the transients and accidents for CANDU reactors. Nevertheless, it is indicated that there are some works to be resolved, such as modeling of headers and multi-channel simulation for the reactor core, etc. Therefore, this study has been initiated with an aim to identify the code applicability for all the postulated transients and accidents in CANDU reactors. In the present study, the small inlet header break experiment (B8604) in the RD-14 test facility was simulated with RELAP5/MOD3.2 code. The RELAP5 results were also compared with both experimental data and those of CATHENA analyses performed by AECL and the analyses demonstrated the code's capability to predict major . phenomena occurring in the transient with sufficient accuracy for both Qualitative and quantitative viewpoint However, some discrepancies in the depressurization of the primary heat transport system after the break and the consequent time delay of the major phenomena were also observed.

• Proceedings of the KIEE Conference
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• 1999.07a
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• pp.372-375
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• 1999

This paper presents an arc(hot-gas flow field) analysis method in GCB. This method includes the Lorentz's force due to magnetic field, turbulent viscous effect and radiation heat transfer which are indispensable to the analysis of hot-gas flow. To verify the applicability of the Proposed method, steady state hot-Eas flow analysis within a simplified interrupter has been carried out. Inlet boundary pressure values were assumed to be 9.0atm and 12.0atm. For each inlet boundary condition, three cases of hot-gas flow field analyses were performed according to the values of arc currents which were assumed to be D.C 0.6kA. 1.0kA and 2.0kA. The results revealed that the arc radius at nozzle throat has been concentrated by increasing the pressure of nozzle upstream and that the maximum temperature of arc core has been decreased along to nozzle exit and the high temperature lesion come to be wide in nozzle downstream. From these results, it is confirmed that the proposed method will be applicable to predict the large current interruption capability of GCB.

• Journal of the Korean Society of Manufacturing Process Engineers
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• v.18 no.6
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• pp.55-61
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• 2019

The purpose of this study was to numerically analyze the heat flow characteristics of an automotive radiator. Heat flow analyses were conducted on the cooling water and outdoor air of the radiator, as well as the temperature distribution of the cooling water after heat transfer. The results of the study revealed that neither heat transfer nor radiator volume was affected by the position of the inlet of cooling water. However, temperature distribution was affected by the position of both the inlet and outlet. In case of heat transfer, three models underwent about 158 kW of heat transfer. The difference in cooling water temperature was about $10^{\circ}C$. In case of pressure drop, the core external air side was reduced to about 1,375 Pa, and the internal cooling water side about 14,570 Pa.

• Nuclear Engineering and Technology
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• v.18 no.3
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• pp.209-217
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• 1986

The inlet coolant temperature reduction technique has been used to extend the load follow operation further in the end-of-cycle-life(EOL). In order to simulate the technique and calculate the nuclear characteristics of a PWR core according to the load follow operation, the three dimensional computing system has been established. The analysis was performed in both MINB and SPINR modes of typical 12-3-6-3 locad follow operation for the EOL of KNU-1 plant. Moreover, the capability of return-to-power has been also tested for those two modes with the system analysis by the RETRAN-02 code. The results show that it has no difficulty to extend the load follow operation further in the EOL by applying the inlet coolant temprature reduction, and also the spinning reserve capacity(SRC) increases by 13% in MINB mode and 14% in SPINR mode Bore that used by control rods only, for 14$^{\circ}$ F drop in the inlet temperature.

• Journal of the Korean Society of Propulsion Engineers
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• v.21 no.3
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• pp.49-55
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• 2017

In this study, gas turbine engine inlet duct was designed to satisfy uniform flow at aerodynamic interface plane (AIP). Haack-series was selected as nose cone profile and duct outer radius($r_o$) was designed to satisfy to match with area change rate between the nose cone and outer duct wall by the 1-D sizing. The design object of the inlet duct wall profile which has the gradual area change rate was uniform Mach number in the core flow region and minimum boundary later thickness at the both inner nose wall and outer duct wall. The flow characteristics inside the inlet duct was evaluated using CFD. The static pressure distribution at the AIP showed uniform pattern within 0.16%. Based on Mach number profile, the boundary layer thickness was 2% of channel height. Kiel temperature rake location was decided less than 100 mm in front of nose cone where the Mach number is less than 0.1 in order to maximize the temperature probe recovery rate.

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.363-374
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• 2005

A heterogeneous thorium-based Kyung Hee Thorium Fuel (KTF) assembly design was assessed for application in the APR-1400 to study the feasibility of using thorium fuel in a conventional pressurized water reactor (PWR). Thermal hydraulic safety was examined for the thorium-based APR-1400 core, focusing on the Departure from Nucleate Boiling Ratio (DNBR) and Large Break Loss of Coolant Accident (LBLOCA) analysis. To satisfy the minimum DNBR (MDNBR) safety limit condition, MDNBR>1.3, a new grid design was adopted, that enabled grids in the seed and blanket assemblies to have different loss coefficients to the coolant flow. The fuel radius of the blanket was enlarged to increase the mass flow rate in the seed channel. Under transient conditions, the MDNBR values for the Beginning of Cycle (BOC), Middle of Cycle (MOC), and End of Cycle (EOC) were 1.367, 1.465, and 1.554, respectively, despite the high power tilt across the seed and blanket. Anticipated transient for the DNBR analysis were simulated at conditions of $112\%$ over-power, $95\%$ flow rate, and $2^{\circ}C$ higher inlet temperature. The maximum peak cladding temperature (PCT) was 1,173K for the severe accident condition of the LBLOCA, while the limit condition was 1,477K. The proliferation resistance potential of the thorium-based core was found to be much higher than that of the conventional $UO_2$ fuel core, $25\%$ larger in Bare Critical Mass (BCM), $60\%$ larger in Spontaneous Neutron Source (SNS), and $155\%$ larger in Thermal Generation (TG) rate; however, the radio-toxicity of the spent fuel was higher than that of $UO_2$ fuel, making it more environmentally unfriendly due to its high burnup rate.

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.452-459
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• 2023

The core power control is an important issue for the study of dynamic characteristics in China initiative accelerator driven subcritical system (CiADS), which has direct impact on the control strategy and safety analysis process. The CiADS is an experimental facility that is only controlled by the proton beam intensity without considering the control rods in the current engineering design stage. In order to get the optimized operation scheme with the stable and reliable features, the variation of beam intensity using the continuous and periodic control approaches has been adopted, and the change of collimator and the adjusting of duty ratio have been proposed in the power control process. Considering the neutronics and the thermal-hydraulics characteristics in CiADS, the physical model for the core power control has been established by means of the point reactor kinetics method and the lumped parameter method. Moreover, the multi-inputs single-output (MISO) logical structure for the power control process has been constructed using proportional integral derivative (PID) controller, and the meta-heuristic algorithm has been employed to obtain the global optimized parameters for the stable running mode without producing large perturbations. Finally, the verification and validation of the control method have been tested based on the reference scenarios in considering the disturbances of spallation neutron source and inlet temperature respectively, where all the numerical results reveal that the optimization method has satisfactory performance in the CiADS core power control scenarios.

• Journal of Ceramic Processing Research
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• v.13 no.spc2
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• pp.328-331
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• 2012

With the recent development of super-precision optical instruments, camera modules for devices, such as portable terminals and digital camera lenses, are increasingly being used. Since an optical lens is usually produced by high-temperature compression molding methods using tungsten carbide (WC) alloy molding cores, it is necessary to develop and study technology for super-precision processing of molding cores and coatings for the core surface. In this study, Rhenium-Iridium (Re-Ir) thin films were deposited onto a WC molding core using a sputtering system. The Re-Ir thin films were prepared by a multi-target sputtering technique, using iridium, rhenium, and chromium as the sources. Argon and nitrogen were introduced through an inlet into the chamber to be the plasma and reactive gases. The Re-Ir thin films were prepared with targets having a composition ratio of 30 : 70, and the Re-Ir thin films were formed with a 240 nm thickness. Re-Ir thin films on WC molding core were analyzed by scanning electron microscope (SEM), atomic force microscope (AFM), and Ra (the arithmetical average surface roughness). Also, adhesion strength and coefficient friction of Re-Ir thin films were examined. The Re-Ir coating technique has received intensive attention in the coating processes field because of promising features, such as hardness, high elasticity, abrasion resistance and mechanical stability that result from the process. Re-Ir coating technique has also been applied widely in industrial and biomedical applications. In this study, WC molding core was manufactured, using high-performance precision machining and the effects of the Re-Ir coating on the surface roughness.