• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2004년도 추계 학술대회논문집
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• pp.151-154
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• 2004

The HANARO, a multi-purpose research reactor of 30 MWth open-tank-in-pool type, has been under normal operation since its initial criticality in February, 1995. Many experiments should be safely performed to activate the utilization of the HANARO. A flow simulated test facility has been developed for the verification of structural integrity of those experimental facilities prior to loading In the HANARO. This test facility is composed of three major parts; a half-core structure assembly, flow circulation system and support system. The half-core structure assembly is composed of plenum, grid plate, core channel with flow tubes, chimney and dummy pool. The flow channels are to be filled with flow orifices to simulate similar flow characteristics to the HANARO. This paper describes an analysis of the flow distribution of the cote channel and compares with the test results. As results, the analysis showed similar flow characteristics compared with those in the test results.

• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 2002년도 유체기계 연구개발 발표회 논문집
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• pp.419-424
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• 2002

The HANARO, a multi-purpose research reactor of 30 MWth open-tank-in-pool type, has been under normal operation since its initial criticality In February, 1995. Many experiments should be safely performed to activate the utilization of the HANARO. A flow simulation facility is being developed for the endurance test of reactivity control units for extended life times and the verification of structural integrity of those experimental facilities prior to loading in the HANARO. This test facility is composed of three major parts; a half-core structure assembly, flow circulation system and support system. The flow circulation system is composed of a circulation pump, a core flow pipe, a core bypass flow pipe and instruments. The system is to be filled with de-mineralized water and the flow should be met the design flow to simulate similar flow characteristics in the core channel of the half-core test facility to the HANARO. This paper, therefore, describes an analytical analysis to study the flow behavior of the system. The computational flow analysis has been performed for the verification of system pressure variation through the three-dimensional analysis program with standard k-$\epsilon$ turbulence model and for the verification of the structural piping integrity through the finite element method. The results of the analysis are satisfied the design requirements and structural piping integrity of flow circulation system.

• Nuclear Engineering and Technology
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• 제45권3호
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• pp.347-360
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• 2013

This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

• Wind and Structures
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• 제12권3호
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• pp.219-237
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• 2009

This study examines the accuracy of large-eddy simulation (LES) to simulate the flow around a large irregular sloping complex terrain. Typically, real built up environments are surrounded by complex terrain geometries with many features. The complex terrain surrounding The Hong Kong University of Science and Technology campus was modelled and the flow over an uphill slope was simulated. The simulated results, including mean velocity profiles and turbulence intensities, were compared with the flow characteristics measured in a wind tunnel model test. Given the size of the domain and the corresponding constraints on the resolution of the simulation, the mean velocity components within the boundary layer flow, especially in the stream-wise direction were found to be reasonably well replicated by the LES. The turbulence intensity values were found to differ from the wind tunnel results in the building recirculation zones, mostly due to the constraints placed on spatial and temporal resolutions. Based on the validated mean velocity profile results, the flow-structure interactions around these buildings and the surrounding terrain were examined.

• 한국유체기계학회 논문집
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• 제15권4호
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• pp.5-11
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• 2012

To identify the flow characteristics of the SMART reactor, a flow distribution model test and a numerical simulation are performed in KAERI. Among several part of the SMART reactor, the fuel assemblies are simulated using simulators because of the complexity. The geometries of the core in the SMART reactor and simulator are different, but some similarities are maintained such as the ratio of pressure drop in the vertical and cross directions. There are cross flow holes in each core simulator to reproduce the cross flow of SMART fuel assemblies. To know the flow characteristics of the cross flow, numerical analysis is performed. As the cross flow area is decreased, the pressure drop between inlet and outlet is decreased. Also, when the flow imbalance between two core simulators is constant, the cross flow area does not significantly affect the cross flow.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3892-3901
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• 2021

The pressure drop of a moisture separator in a steam generator is the important design parameter to ensure the successful performance of a nuclear power plant. The moisture separators have a wide range of operating conditions based on the arrangement of them. The prediction of the pressure drop in a moisture separator is challenging due to the complexity of the multi-dimensional two-phase vortex flow. In this study, the moisture separator test facility using the air/water two-phase flow was used to predict the pressure drop of a moisture separator in a Korean OPR-1000 reactor. The prototypical steam/water two-phase flow conditions in a steam generator were simulated as air/water two-phase flow conditions by preserving the centrifugal force and vapor quality. A series of experiments were carried out to investigate the effect of hydraulic characteristics such as the quality and liquid mass flux on the two-phase pressure drop. A new prediction model based on the scaling law was suggested and validated experimentally using the full and half scale of separators. The suggested prediction model showed good agreement with the steam/water experimental results, and it can be extended to predict the steam/water two-phase pressure drop for moisture separators.

• Corrosion Science and Technology
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• 제15권5호
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• pp.245-252
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• 2016

The pipelines and equipments are degraded by flow-accelerated corrosion (FAC), and a large-scale test facility was constructed for simulate the FAC phenomena in secondary coolant environment of PWR type nuclear power plants. Using this facility, FAC test was performed on weld pipe (carbon steel and low alloy steel) at the conditions of high velocity flow (> 10 m/s). Wall thickness was measured by high temperature ultrasonic monitoring systems (four-channel buffer rod type and waveguide type) during test period and room temperature manual ultrasonic method before and after test period. This work deals with the complex effects of flow velocity on the wall thinning in weld pipe and the test results showed that the higher flow velocity induced different increasement of wall thinning rate for the carbon steel and low alloy steel pipe.

• Nuclear Engineering and Technology
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• 제50권8호
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• pp.1412-1420
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• 2018

An experiment was performed using the large-scale test facility (LSTF), which simulated a 1% vessel upper head small-break loss-of-coolant accident with an accident management (AM) measure under an assumption of total-failure of high-pressure injection (HPI) system in a pressurized water reactor (PWR). In the LSTF test, liquid level in the upper head affected break flow rate. Coolant was manually injected from the HPI system into cold legs as the AM measure when the maximum core exit temperature reached 623 K. The cladding surface temperature largely increased due to late and slow response of the core exit thermocouples. The AM measure was confirmed to be effective for the core cooling. The RELAP5/MOD3.3 code indicated insufficient prediction of primary coolant distribution. The author conducted uncertainty analysis for the LSTF test employing created phenomena identification and ranking table for each component. The author clarified that peak cladding temperature was largely dependent on the combination of multiple uncertain parameters within the defined uncertain ranges.

• Nuclear Engineering and Technology
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• 제41권10호
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• pp.1263-1274
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• 2009

Several experimental tests to simulate a reflood phase of a cold-leg LBLOCA of the APR1400 have been performed using the ATLAS facility. This paper describes the related experimental results with respect to the thermal-hydraulic behavior in the core and the system-core interactions during the reflood phase of the cold-leg LBLOCA conditions. The present descriptions will be focused on the LB-CL-09, LB-CL-11, LB-CL-14, and LB-CL-15 tests performed using the ATLAS. The LB-CL-09 is an integral effect test with conservative boundary condition; the LB-CL-11 and -14 are integral effect tests with realistic boundary conditions, and the LB-CL-15 is a separated effect test. The objectives of these tests are to investigate the thermal-hydraulic behavior during an entire reflood phase and to provide reliable experimental data for validating the LBLOCA analysis methodology for the APR1400. The initial and boundary conditions were obtained by applying scaling ratios to the MARS simulation results for the LBLOCA scenario of the APR1400. The ECC water flow rate from the safety injection tanks and the decay heat were simulated from the start of the reflood phase. The simulated core power was controlled to be 1.2 times that of the ANS-73 decay heat curve for LB-CL-09 and 1.02 times that of the ANS-79 decay curve for LB-CL-11, -14, and -15. The simulated ECC water flow rate from the high pressure safety injection pump was 0.32 kg/s. The present experimental data showed that the cladding temperature behavior is closely related to the collapsed water level in the core and the downcomer.

• 한국추진공학회:학술대회논문집
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• 한국추진공학회 2011년도 제36회 춘계학술대회논문집
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• pp.144-152
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• 2011

고고도에서 장시간 운용되는 무인항공기용 추진시스템의 고고도 운용 특성을 연구하기 위하여 고도엔진시험설비를 이용한 엔진성능 시험이 필요하다. 기존의 범용 고도엔진시험설비를 이용하려면 시험설비와 엔진 사이에 부가적인 시험설비가 요구되는데, 이중 엔진 성능에 직접 영향을 미치는 흡입구 및 배기덕트의 적절한 설계가 매우 중요하다. 만일 흡입구 및 배기덕트가 적절히 설계되지 않을 경우 유동특성 변화와 압력손실 저하로 인한 엔진 성능 저하가 실제 비행 상태와 달라질 수 있기 때문이다. 본 연구에서는 범용 고도시험설비 내 엔진 흡입구 및 배기덕트의 형상을 3D 모델링하고 ANSYS사의 CFX 전산유체역학(CFX) 프로그램을 이용하여 설계된 형상의 유동 및 압력손실 해석을 수행한 다음 엔진성능해석 프로그램을 이용하여 덕트 손실을 고려한 엔진에 성능을 분석한다. 마지막으로 반복적인 설계형상 해석을 통해 최적화된 흡입구 및 배기덕트 형상을 제안한다.