• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 추계학술대회논문집B
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• pp.3-8
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• 2001

For the successful design of nuclear reactor, it is very important to investigate thermal-hydraulic characteristics of fuel rod bundle. Fluid flow and heat transfer in the non-circular cross-section of nuclear fuel rod bundle are different from those found in common circular tube. And complex three dimensional flow including secondary and vortex flow, is formed around the bundles. The purpose of this research is to examine how geometries and flow conditions affect heat transfer in fuel rod bundle. Design data for nuclear fuel rod bundle and structure are surveyed, and $3{\times}3$ sub-channel model is adopted in this study. Computational results are compared with the heat transfer data measured by naphthalene sublimation method, and numerical analysis and evaluation are performed at various design conditions and flow conditions.

• 유체기계공업학회:학술대회논문집
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• 유체기계공업학회 2004년도 유체기계 연구개발 발표회 논문집
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• pp.473-478
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• 2004

This work presents the Thermal Hydraulic analysis has been performed for a 19-pin wire-spacer fuel assembly using three-dimensional Reynolds-averaged Navier-Stokes equations. SST model is used as a turbulence closure. The whole fuel assembly has been analyzed for one period of the wire-spacer using periodic boundary condition at inlet and outlet of the calculation domain. The overall results far a preliminary calculation show a good agreement with the experimental observations. It has been found that the major unidirectional flows are the axial velocity in sub-channels and the peripheral sweeping flows and the velocities are found to be following a cyclic path of period equal to the wire-wrap pitch. The temperature is found to be maximum in the central region and also, there exist a radial temperature gradient between the fuel rods. The major advantage of performing this kind of analysis is the prediction of thermal-hydraulic behavior of a fuel assembly with much ease.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 춘계학술발표회논문집(1)
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• pp.666-671
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• 1998

Computational fluid dynamics(CFD) code FLUENT[11 was used to simulate the thermal hydraulic processes occuring in conceptual design of the accelerator-driven subcritical reactor(ADSR) liquid lead target. The purpose of the analysis is to investigate the thermal hydraulic characteristics of liquid lead as ADSR target material with various target geometries and injection locations of proton beam. In the calculation analysis, the local temperature of the liquid lead target rises to the boiling temperature very rapidly When the proton beam is injected from the bottom of the target system, the duration time to reach the boiling temperature is longer and the temperature distribution is flatter than other cases.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.639-644
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• 1997

The work reported in this paper identifies the thermal-hydraulic phenomena that are expected to occur during a number of key transients in SMART(System-integrated Modular Advanced ReacTor) which is under development at KAERI. The result of this effort is based on the current design concept of SMART integral reactor. Although the design is still evolving, the preliminary phenomena Identification and Ranking Table(PIRT) has been developed based on the experts' knowledge and experience. The preliminary PIRT has been developed by consensus of KAERI expert panelists and AHP(Analytical Hierarchy Process). Preliminary PIRT developed in this paper is intended to be used to identify and integrate development areas of further experimental tests needed, thermal hydraulic models and correlations and code improvements for the safety analysis of the SMART.

• 한국전산유체공학회:학술대회논문집
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• 한국전산유체공학회 2002년도 추계 학술대회논문집
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• pp.64-69
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• 2002

A numerical study has been peformed for evaluation of convection schemes for thermal hydraulic analysis in a liquid metal reactor Four convection schemes, HYBRID, QUICK, SMART and HLPA included in the CFX-4 code are considered. The performances of convection schemes are evaluated by applying them to the five test problems. The accuracy, stability and convergence are tested. It is shown that the HYBRID scheme is too diffusive, and the QUICK scheme exhibits overshoots and undershoots, and the SMART scheme shows convergence oscillations, and the HLPA scheme preserves the boundedness without causing convergence oscillations. The accuracies of SMART, QUICK and HLPA schemes are comparable. Thus, the use of HLPA scheme is highly recommended for thermal hydraulic analysis in a liquid metal reactor.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2262-2275
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• 2022

The Korea Atomic Energy Research Institute (KAERI) studied the sodium-water reaction (SWR) minimized steam generator for the safety of the sodium-cooled fast reactor (SFR), and selected the copper bonded steam generator (CBSG) as the optimal concept. This paper introduces the conceptual design of the CBSG and the development of the CBSG sizing analyzer (CBSGSA). The CBSG consists of multiple heat transfer modules with a crossflow heat transfer configuration where sodium flows horizontally and water flows vertically. The heat transfer modules are stacked along a vertical direction to achieve the targeted large heat transfer capacity. The CBSGSA code was developed for the thermal-hydraulic analysis of the CBSG in a multi-pass crossflow heat transfer configuration. Finally, we conducted a preliminary sizing and rating analysis of the CBSG for the trans-uranium (TRU) core system using the CBSGSA code proposed by KAERI.

• 한국지반공학회논문집
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• 제19권5호
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• pp.59-67
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• 2003

본 연구에서는 지하수 유동 특성 해석의 신뢰도 증진을 위해서 우리나라 심부결정질 암석의 접촉 면적과 간극의 변화양상, 구속압과 온도에 따라 유체투과율의 변화 양상을 살펴보고자 하였다. 또한 역학적 물성과 열적 물성을 측정하여 모델링의 입력자료에 대한 신빙성을 높일 수 있도록 하였다. 실험 결과 초기 평균 간극은 544.33$\mu\textrm{m}$∼898.62$\mu\textrm{m}$ 범위를 보였으며 분포특성은 로그-정규분포 형태를 보였다. 구속압 변화에 따라 유체투과율은 감소하는 경향이 나타나고 온도 증가에 따라 유체투과율이 감소하는 경향을 보였으며 접촉 면적이 커지면 유체투과율 변화 폭이 작아짐을 이론과 실험을 통해 규명하였다. 암석의 역학적, 열적 물성을 측정한 결과 기존에 측정된 우리나라 화강암의 물성치와 근사한 값을 가지는 것으로 나타났다.

• Nuclear Engineering and Technology
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• 제45권7호
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• pp.871-894
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• 2013

KAERI (Korea Atomic Energy Research Institute) has been operating an integral effect test facility, the Advanced Thermal-Hydraulic Test Loop for Accident Simulation (ATLAS), for transient and accident simulations of advanced pressurized water reactors (PWRs). Using ATLAS, a high-quality integral effect test database has been established for major design basis accidents of the APR1400 plant. A Domestic Standard Problem (DSP) exercise using the ATLAS database was promoted to transfer the database to domestic nuclear industries and contribute to improving a safety analysis methodology for PWRs. This $2^{nd}$ ATLAS DSP (DSP-02) exercise aims at an effective utilization of an integral effect database obtained from ATLAS, the establishment of a cooperation framework among the domestic nuclear industry, a better understanding of the thermal hydraulic phenomena, and an investigation into the possible limitation of the existing best-estimate safety analysis codes. A small break loss of coolant accident with a 6-inch break at the cold leg was determined as a target scenario by considering its technical importance and by incorporating interests from participants. This DSP exercise was performed in an open calculation environment where the integral effect test data was open to participants prior to the code calculations. This paper includes major information of the DSP-02 exercise as well as comparison results between the calculations and the experimental data.

• 드라이브 ㆍ 컨트롤
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• 제18권3호
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• pp.8-15
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• 2021

A rear axle steering (RAS) system is attached to the rear of medium and large commercial vehicles that transport large cargo. The existing RAS systems are driven by electro-hydraulic actuator (EHA), and most commercialized EHAs consist of electric motors, hydraulic pumps, relief valves, prefill valves and cylinders. The prefill valve required for such EHAs is a type of check valve with extremely low cracking pressure that should not allow RAS to have noise or vibration, and the prefill valve prevents system negative pressure as well as unstable operation. Most papers on this topic rely on experiments to predict valve performance, and theoretically detailed modeling of valves or pipelines is performed, but it is very rare to evaluate hydraulic vibration characteristics by analysing everything from hydraulic pumps to valves comprehensively. In this study, we proposed an experimental circuit that can predict the performance of the prefill valve. The study also analysed the pressure-flow pulsation that is transmitted to the valve through the pipeline, and how the transmitted pressure-flow pulsation affects the valve vibration.

• 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.