• 한국산학기술학회논문지
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• 제16권5호
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• pp.3030-3034
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• 2015

새로운 엔진과 차량을 개발하여 엔진을 차량에 탑재할 때, 중요하게 고려해야 할 사항 중의 하나는 냉각 성능이다. 만약 냉각 성능이 열악하다면 엔진은 과열되어 파손되게 된다. 그러나 자동차회사에서 일반적으로 엔진은 차량보다 훨씬 빠른 시기에 개발이 진행되게 되어 엔진을 차량에 탑재한 조건에서 냉각 성능을 시험할 수 없다. 본 연구에서는 몇 가지 시험과 계산 결과를 이용하여 엔진의 냉각 성능을 추정하였다. 첫 번째로 엔진의 열정산 시험이 진행되었다. 두 번째로 냉각수 유동 시험이 진행되었다. 이 시험에서 라디에이터로 유입되는 유량을 구할 수 있다. 그리고 차량의 냉각 시험 성능 조건으로부터 차량의 부하와 속도를 구하고, 이로부터 엔진의 토크와 rpm이 계산되었다. 그리고 이러한 결과를 비교하여 엔진의 냉각 성능이 추정되었다.

• Nuclear Engineering and Technology
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• 제50권3호
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• pp.368-378
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• 2018

Understanding of turbulent flow in the reactor coolant pump (RCP) is a premise of the optimal design of the RCP. Flow structures in the RCP, in view of the specially devised spherical casing, are more complicated than those associated with conventional pumps. Hitherto, knowledge of the flow characteristics of the RCP has been far from sufficient. Research into the nonintrusive measurement of the internal flow of the RCP has rarely been reported. In the present study, flow measurement using particle image velocimetry is implemented to reveal flow features of the RCP model. Velocity and vorticity distributions in the diffuser and spherical casing are obtained. The results illuminate the complexity of the flows in the RCP. Near the lower end of the discharge nozzle, three-dimensional swirling flows and flow separation are evident. In the diffuser, the imparity of the velocity profile with respect to different axial cross sections is verified, and the velocity increases gradually from the shroud to the hub. In the casing, velocity distribution is nonuniform over the circumferential direction. Vortices shed consistently from the diffuser blade trailing edge. The experimental results lend sound support for the optimal design of the RCP and provide validation of relevant numerical algorithms.

• Nuclear Engineering and Technology
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• 제50권6호
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• pp.829-841
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• 2018

An experiment was conducted for the OECD/NEA ROSA-2 Project using the large-scale test facility (LSTF), which simulated a 17% hot leg intermediate-break loss-of-coolant accident in a pressurized water reactor (PWR). In the LSTF test, core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on the upper core plate in the upper plenum. Results of the uncertainty analysis with RELAP5/MOD3.3 code clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges. For studying the scaling problems to extrapolate thermal-hydraulic phenomena observed in scaled-down facilities, an experiment was performed for the OECD/NEA PKL-3 Project with the Primarkreislaufe Versuchsanlage (PKL), as a counterpart to a previous LSTF test. The LSTF test simulated a PWR 1% hot leg small-break loss-of-coolant accident with steam generator secondary-side depressurization as an accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for the primary pressure, the core collapsed liquid level, and the cladding surface temperature probably due to effects of differences between the LSTF and the PKL in configuration, geometry, and volumetric size.

• 한국산업융합학회 논문집
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• 제5권4호
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• pp.379-384
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• 2002

As automotive diesel engines adopt the direct injection method for a lower level of the exhaust emission and a higher fuel efficiency, the maximum temperature of engine coolant decreases. Consequently, the total available heat source from the engine coolant decreases over 35%. However, the heating source of air-conditioning system in automobiles depends on the hot engine coolant completely, so that it is nearly impossible to control air conditioning in heating season. Therefore, the present study has been carried out to develop the air conditioning system for the high efficient heat pump type using the HFC-134a. Especially, the air conditioning system of heating has been developed at a beginning stage, when it has low heat source from small and medium sized diesel recreation vehicles. To develop a control logic system for air conditioning system which is a heat pump type with a heat recovery exchanger, its cycle characteristics has been investigated according to the opening of LEV at a bench system.

• 한국표면공학회지
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• 제51권4호
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• pp.218-223
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• 2018

In this study, we proposed a new ultrasonic-chemical decontamination process for decontaminating radioactive corrosion products during the maintenance of reactor coolant pump (RCP). The actual decontamination process was reproduced in the laboratory. And the corrosion characteristics of stainless steel (STS), constituting the RCP interior parts, were examined. The weight-loss measurment and polarization experiment were carried out in order to determine the corrosion characteristics of STS 304 and STS 415 by repeated decontamination processes. The STS 304 presented a little corrosion damage, which was almost indistinguishable from visual observation. The weight-loss rate of STS 304 was also significantly lower. On the other hand, STS 415 showed severe corrosion damage on its surface, greater weight-loss rate and higher corrosion current density than STS 304.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2008년 영문 학술대회
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• pp.346-357
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• 2008

Endothermic fuels are known as a probable fuel for hypersonic atmospheric flight vehicles and advanced propulsion systems, as well as cryogenic fuels. Especially, from the standpoint of the advanced regenerative cooling use, they are quite useful as a coolant fuel because of their large heat sink due to their chemical decompositions; so-called endothermic cooling effect. However, no heat transfer equations have been proposed taking into account such endothermic reactive behaviors concretely. This paper describes an analytical method for evaluation of the heat transfer rates between endothermic reacting coolant fuel and coolant-side wall in the regenerative cooling passages. Heat transfer mechanism is indicated based on a classical transport-phenomenological approach. A new relational expression of Nusselt number ratio for forcedconvective heat transfer with such endothermic reactions is also proposed by theoretical approaches using some classical hypotheses. Its applicability is assessed provisionally by comparison with confirmed results of heated tube tests for supercritical JP-7 fuel carried out at NASA Lewis Research Center, using its heat sink characteristics evaluated by United Technologies Research Center(UTRC). As a result, it has been suggested that the proposed relational equation is applicable to the evaluation of enhancement of Nusselt numbers due to such reactions in developed turbulent flows such as in the regenerative cooling passages.

• Nuclear Engineering and Technology
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• 제39권4호
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• pp.327-336
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• 2007

The influence of density differences on the mixing of the primary loop inventory and the Emergency Core Cooling (ECC) water in the downcomer of a Pressurised Water Reactor (PWR) was analyzed at the ROssendorf COolant Mixing (ROCOM) test facility. ROCOM is a 1:5 scaled model of a German PWR, and has been designed for coolant mixing studies. It is equipped with advanced instrumentation, which delivers high-resolution information for temperature or boron concentration fields. This paper presents a ROCOM experiment in which water with higher density was injected into a cold leg of the reactor model. Wire-mesh sensors measuring the tracer concentration were installed in the cold leg and upper and lower part of the downcomer. The experiment was run with 5% of the design flow rate in one loop and 10% density difference between the ECC and loop water especially for the validation of the Computational Fluid Dynamics (CFD) software ANSYS CFX. A mesh with two million control volumes was used for the calculations. The effects of turbulence on the mean flow were modelled with a Reynolds stress turbulence model. The results of the experiment and of the numerical calculations show that mixing is dominated by buoyancy effects: At higher mass flow rates (close to nominal conditions) the injected slug propagates in the circumferential direction around the core barrel. Buoyancy effects reduce this circumferential propagation. Therefore, density effects play an important role during natural convection with ECC injection in PWRs. ANSYS CFX was able to predict the observed flow patterns and mixing phenomena quite well.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.928-940
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• 2017

An experiment using the $Prim{\ddot{a}}rkreisl{\ddot{a}}ufe$ Versuchsanlage (PKL) was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with the large-scale test facility (LSTF) on a cold leg smallbreak loss-of-coolant accident with an accident management (AM) measure in a pressurized water reactor. Concerning the AM measure, the rate of steam generator (SG) secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition; however, the onset timings of the SG depressurization were different from each other. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing, which caused whole core quench. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code predicted the overall trends of the major thermal-hydraulic responses observed in the LSTF test well, and indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.979-988
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• 2017

The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

• Nuclear Engineering and Technology
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• 제53권10호
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• pp.3264-3274
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• 2021

In a severe accident of light water reactor (LWR), molten core material (corium) can be released into the wet cavity, and a fuel-coolant interaction (FCI) can occur. The molten jet with high speed is broken and fragmented into small debris, which may cause a steam explosion or a molten core concrete interaction (MCCI). Since the premixing stage where the jet breakup occurs has a large impact on the severe accident progression, the understanding and evaluation of the jet breakup phenomenon are highly important. Therefore, in this study, the jet breakup simulations were performed using the Smoothed Particle Hydrodynamics (SPH) method which is a particle-based Lagrangian numerical method. For the multi-fluid system, the normalized density approach and improved surface tension model (CSF) were applied to the in-house SPH code (single GPU-based SOPHIA code) to improve the calculation accuracy at the interface of fluids. The jet breakup simulations were conducted in two cases: (1) jet breakup without structures, and (2) jet breakup with structures (control rod guide tubes). The penetration depth of the jet and jet breakup length were compared with those of the reference experiments, and these SPH simulation results are qualitatively and quantitatively consistent with the experiments.