• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.842-848
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• 2022

Parametric studies of heat transfer and fluid flow are very important research of interest because the design and operation of fluid flow and heat transfer systems are guided by these parametric studies. The safety of the system operation and system optimization can be determined by decreasing or increasing particular fluid flow and heat transfer parameter while keeping other parameters constant. The parameters that can be varied in order to determine safe and optimized system include system pressure, mass flow rate, heat flux and coolant inlet temperature among other parameters. The fluid flow and heat transfer systems can also be enhanced by the presence of or without the presence of particular effects including gravity effect among others. The advanced Generation IV reactors to be deployed for large electricity production, have proven to be more thermally efficient (approximately 45% thermal efficiency) than the current light water reactors with a thermal efficiency of approximately 33 ℃. SCWR is one of the Generation IV reactors intended for electricity generation. High Performance Light Water Reactor (HPLWR) is a SCWR type which is under consideration in this study. One-eighth of a proposed fuel assembly design for HPLWR consisting of 7 fuel/rod bundles with 9 coolant sub-channels was the geometry considered in this study to examine the effects of system pressure and mass flow rate on wall and fluid temperatures. Gravity effect on wall and fluid temperatures were also examined on this one-eighth fuel assembly geometry. Computational Fluid Dynamics (CFD) code, STAR-CCM+, was used to obtain the results of the numerical simulations. Based on the parametric analysis carried out, sub-channel 4 performed better in terms of heat transfer because temperatures predicted in sub-channel 9 (corner subchannel) were higher than the ones obtained in sub-channel 4 (central sub-channel). The influence of system mass flow rate, pressure and gravity seem similar in both sub-channels 4 and 9 with temperature distributions higher in sub-channel 9 than in sub-channel 4. In most of the cases considered, temperature distributions (for both fluid and wall) obtained at 25 MPa are higher than those obtained at 23 MPa, temperature distributions obtained at 601.2 kg/h are higher than those obtained at 561.2 kg/h, and temperature distributions obtained without gravity effect are higher than those obtained with gravity effect. The results show that effects of system pressure, mass flowrate and gravity on fluid flow and heat transfer are significant and therefore parametric studies need to be performed to determine safe and optimum operating conditions of fluid flow and heat transfer systems.

• Nuclear Engineering and Technology
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• v.36 no.3
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• pp.248-262
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• 2004

Turbulence models are separately assessed for a three dimensional thermal-hydraulic analysis of the integral reactor SMART. Seven models (mixing length, k-l, standard $k-{\epsilon},\;k-{\epsilon}-f{\mu},\;k-{\epsilon}-v2$, RRSM, and ERRSM) are investigated for flat plate channel flow, rotating channel flow, and square sectioned U-bend duct flow. The results of these models are compared to the DNS data and experiment data. The results are assessed in terms of many aspects such as economical efficiency, accuracy, theorization, and applicability. The standard $k-{\epsilon}$ model (high Reynolds model), the $k-{\epsilon}-v2$ model, and the ERRSM (low Reynolds models) are selected from the assessment results. The standard $k-{\epsilon}$ model using small grid numbers predicts the channel flow with higher accuracy in comparison with the other eddy viscosity models in the logarithmic layer. The elliptic-relaxation type models, $k-{\epsilon}-v2$, and ERRSM have the advantage of application to complex geometries and show good prediction for near wall flows.

• The KSFM Journal of Fluid Machinery
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• v.19 no.4
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• pp.19-28
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• 2016

Objective of this study is to compare the inherent characteristics of natural convection flow inside the canister of spent fuel dry storage system with different backfill gases by utilizing computational fluid dynamics (CFD) code. Four working fluids were selected for comparison study. Helium currently used backfill gas for canister, air, nitrogen, and argon are frequently used as coolant in many heat transfer applications. The results indicate that helium has very distinct conductive behavior and show very weak natural convective flow compared to the others. Argon showed the strongest natural convective flow but also the worst coolability. Air and nitrogen showed similar characteristics to each other. However, due to difference in Prandtl number, nitrogen showed more effective natural convective flow. These results suggest that experimental validation for the nitrogen is needed to investigate the potential coolability other than currently commercially used helium.

• Proceedings of the Korea Institute of Applied Superconductivity and Cryogenics Conference
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• 2003.02a
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• pp.167-170
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• 2003

The thermal and hydraulic characteristics of the cooling passage for HTS power cable should be carefully investigated to get the highly reliable and economic operation. In this study, the pressure drop and temperature change of the coolant flowing the corrugated passage was estimated using the simple one-dimensional approach, and the temperature distribution in the radial direction was calculated. The results show the dependency of the mass flow rate, thermal conductivities on the cooling performance of the HTS cable.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.3175-3180
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• 2008

A metal-fueled pool-type liquid metal fast reactor (LMFR) provides large margins to sodium boiling and fuel damage under accident conditions. The favorable passive safety results are obtained by both a reactivity feedback mechanism in the core and a passive decay heat removal system. Among the various reactivity feedbacks, the ones by a thermal expansion of a radial dimension of the core and by the control rod drivelines are strongly dependent on the flow conditions in the core and the hot pool, respectively. The effects of multidimensional thermal hydraulic characteristics on these reactivity feedbacks are investigated by the system-wide safety analysis code SSC-K with advanced thermal hydraulics models. Particularly a detailed three dimensional thermal hydraulics reactor core model is integrated into SSC-K for use in a whole system analysis of the passive safety aspects of LMR designs. The model provides fuel and cladding temperatures for every fuel pin in a reactor and coolant temperatures for every coolant sub-channel in the reactor.

• Nuclear Engineering and Technology
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• v.44 no.8
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• pp.831-846
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• 2012

KAERI has developed a two-phase CFD code, CUPID, for a refined calculation of transient two-phase flows related to nuclear reactor thermal hydraulics, and its numerical models have been verified in previous studies. In this paper, the CUPID code is validated against experiments on the downcomer boiling and moderator flow in a Calandria vessel. Physical models relevant to the validation are discussed. Thereafter, multi-scale thermal hydraulic analyses using the CUPID code are introduced. At first, a component-scale calculation for the passive condensate cooling tank (PCCT) of the PASCAL experiment is linked to the CFD-scale calculation for local boiling heat transfer outside the heat exchanger tube. Next, the Rossendorf coolant mixing (ROCOM) test is analyzed by using the CUPID code, which is implicitly coupled with a system-scale code, MARS.

• Journal of Advanced Marine Engineering and Technology
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• v.22 no.6
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• pp.771-781
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• 1998

A highly compact and thermally efficient water cooled oil cooler for automotive use without offset strip fin and casing is developed in this study. The study result has shown that eliminating the fin and casing in the oil cooler the manufacturing process and cost and can be simplified and reduced greatly without sacrificing the thermal capacity. The oil cooler developed in the study uses the dimply type heat transfer core element design instead of offset strip fin and eliminates the outer casing for coolant water flow by applying specially made parallel loop flow design. In the study the thermal design program for the present oil cooler also was developed and validated experimentally.

• Transactions of Materials Processing
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• v.20 no.4
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• pp.291-295
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• 2011

The injector housing has two functions, namely, positioning the injector and protecting it from coolant. The conventional manufacturing process of the injector housing by machining has some drawbacks such as considerable loss of material and environmental pollution caused by excessive use of cutting oil. In this paper, precision cold forging is proposed as a new manufacturing process in order to improve these issues. A numerical study was conducted to compute the metal flow, strain, load and other process variables using DEFORM-2D, a finite element analysis(FEA) code for metal forming. Two process methods were investigated and optimal conditions were computed with the FEA code. A prototype was manufactured from the optimal process method and the metal flow and hardness were obtained from the prototype.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2821-2835
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• 2020

The phenomenon identification and ranking table (PIRT) is an important basis in the nuclear power plant (NPP) thermal-hydraulic analysis. This study focuses on the importance ranking of the input parameters when lacking the PIRT, and the target scenario is the small break loss of coolant accident (SBLOCA) in a pressurized water reactor (PWR) CPR1000. A total of 54 input parameters which might have influence on the figure of merit (FOM) were identified, and the sensitivity measure of each input on the FOM was calculated through an optimized moment-independent global sensitivity analysis method. The importance ranking orders of the parameters were transformed into the Savage scores, and the parameters were categorized based on the Savage scores. A parameter importance ranking table for the SBLOCA scenario of the CPR1000 reactor was obtained, and the influences of some important parameters at different break sizes and different accident stages were analyzed.

• Proceedings of the Korean Society of Propulsion Engineers Conference
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• 2017.05a
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• pp.112-115
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• 2017

We analyzed the rocket engine flow to check whether the possibility of the ground test and the equipment safety problems in the high altitude engine test facility. The test condition is that the vacuum chamber is open and the coolant water is injected into the supersonic diffuser. The analysis uses two-dimensional axisymmetry with a mixture of plume, air, and cooling water. As a result, the ground test was possible up to the cooling water flow rate of 200 kg/sec. However, due to the back flow of the initial plume, the vacuum chamber is exposed to high temperature, and at the same time, the inside of the vacuum chamber is contaminated due to the reverse flow of the cooling water. Therefore, sufficient insulation measures and work for pollution avoidance should be preceded.