• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.209.1-209
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• 2002

• 유체기계공업학회:학술대회논문집
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• 2004.12a
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• pp.546-551
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• 2004

An applying Thermodynamic method for the purpose of measuring hydraulic efficiency of pump-motor system, based on IEC60041 code, is not easy to adopt at field test. Even though there were splendid development in measuring technic in discharge measuring through the hydraulic machina lots of unsolved problems concerned in flow-rate are still remain in measuring hydraulic efficiency in hydraulic machine. The key point in measuring hydraulic efficiency is to measure exact flow-rate. So, Thermodynamic methode provides a good solution. This methode measures hydraulic efficiency by detecting the difference of temperature and pressure between the hydraulic process of machine, without measuring flow-rate of pump or turbine. By measuring temperature in mk level and absolute pressure in pascal, we can get a difference of thermodynamic specific energy in Moliere chart before and after of hydraulic process, md that difference is equal to hydraulic loses. Following the standard in proceeding Thermodynamic methode, I hope these trial and records make others be familiar to the thermal methode and make it easer to beginner for trial.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.391-396
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• 1996

내부부수로, 벽면부수로, 모서리부수로를 포함하는 가압경수로형 원자로의 핵연료집합체를 모의하는 3$\times$3 봉다발을 모델로 수치해석을 통해 봉다발 주변의 유동특성을 알아보고 각 봉에서의 원주방향 위치에 따른 국소열전달 특성에 관해 고찰하였다. 봉다발에서 열전달계수의 분포는 벽면영향으로 인한 각 부수로에서의 유속분포와 밀접한 관계가 있으며 내부부수로에 인접한 봉에서 가장 높았고, 그 다음이 벽면부수로, 모서리부수로에 인접한 봉에서는 가장 작게 나타났다. 현재 핵연료의 열수력 설계시에 적용하고 있는 부수고 내의 모든 열수력학적 변수가 일정하다고 가정하는 부수로 해석방법은 봉다발내의 실제 열전달 현상과는 상당한 차이가 있음을 보여주었다.

• Nuclear Engineering and Technology
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• v.23 no.1
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• pp.56-65
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• 1991

The LOFT LOCE L2-3 was simulated using the RELAP5/MOD2 Cycle 36.04 code to assess its capability in predicting the thermal-hydraulic phenomena in LBLOCA of a PWR. The reactor vessel was simulated with two core channels and split downcomer modeling for a base case calculation using the frozen code. The result of the base calculation showed that the code predicted the hydraulic behavior, and the blowdown thermal response at high power region of the core reasonably and that the code had deficiencies in the critical How model during subcooled-two-phase transition period, in the CHF correlation at high mass flux and in the blowdown rewet criteria. An overprediction of coolant inventory due to the deficiencies yielded the poor prediction of reflood thermal response. Improvement of the code, RELAP5 / MOD2 Cycle 36.04, based on the sensitivity study increased the accuracy of the prediction of the rewet phenomena.

• Nuclear Engineering and Technology
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• v.25 no.2
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• pp.343-352
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• 1993

다목적연구로(KMRR)는 일반 발전용 원자로와는 매우 다른 특성을 가지고 있으며, 설계 개념 또한 특이하다. 위와 같은 특이한 설계 특성을 파악하기 위하여 열수력 실험을 수행하였으며 시운전 시험도 설계 개념의 입증에 중점을 두고 수행될 예정이다. 실증실험은 크게 설계 자료 생산을 위한 실험, 기기 설계 검증 시험, 시운전 성능 시험으로 나눌 수 있다. 설계 자료 생산을 위한 실험으로 핵연료의 열수력학적 특성을 규명하는 실험, 우회 유동에 의한 노심 출구 냉각수 상승 억제를 입증 또는 해석하기 위한 자료 생산용 실험 등이 이루어졌다. 기기 설계 검증 시험으로는 Pump 특성 시험, Flap valve 특성 시험 등을 들 수 있다. 또한, 시운전 성능 시험으로는 설계 개념을 입증하기 위한 여러 시험들이 행해질 예정이다. 이러한 실험들을 통하여 설계에 필요한 많은 자료들이 생산되었고, 시운전 시험을 통하여 설계를 검증하고 실제 운전에 필요한 많은 자료를 얻을 수 있으리라 기대된다. 본 기고를 통하여 이러한 실험의 중요성 및 내용에 대해 간략하게 기술하고자 한다.

• Nuclear Engineering and Technology
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• v.28 no.2
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• pp.162-174
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• 1996

The local hydraulic characteristics in a single phase flow of a 6$\times$6 rod bundle with neighboring different spacer grids were measured by using a LDV(Laser Doppler Velocimeter) system. 6$\times$6 rod bundle is formed by two 3$\times$6 rod bundles with different spacer grids. The objective of this study in a rod bundle is to investigate the thermal-hydraulic interactions between different spacer grids with different configurations and resistance. By using a LDV system, the velocity and turbulent intensity in axial and horizontal directions ore measured. Pressure drop measurements ore also performed to evaluate the loss coefficient for the spacer grid and the friction factor for rod bundles. Implications concerning thermal mining due to spacer grids were investigated based on the hydraulic test results. Swirl factor, which is assumed as a qualitative criteria for DNB(departure from nucleate boiling), was defined and estimated from the horizontal velocity result.

• Journal of Energy Engineering
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• v.26 no.3
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• pp.90-96
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• 2017

When the hot water is supplied through the district heating (DH) pipeline, a pressure differential control valve (PDCV) protects the DH user equipment from the high pressure DH water and helps to supply DH water to long distance. It also controls the temperature and adjust the pressure in the main district heating pipeline. However, cavitation occurs in PDCV due to the use of high pressure DH water. It causes frequent failures and many problems. It also causes energy loss and complaints to both operators and users. In order to solve these problems, we will introduce the energy saving technology to replace the primary side PDCV with hydraulic turbine, convert the differential pressure into electricity, and utilize electricity as the power of the secondary side pump.

• Nuclear Engineering and Technology
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• v.18 no.1
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• pp.9-16
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• 1986

Simulation of the system thermal-hydraulic parameters was carried out following the KNUl (Korea Nuclear Unit-1) loss of normal feedwater transient sequence occurred on November 14, 1984. Results were compared with the plant transient data, and good agreements were obtained. Some deviations were found in the parameters such as the steam flowrate and the RCS (Reactor Coolant System) average temperature, around the time of reactor trip. It can be expected since the thermal-hydraulic parameters encounter rapid transitions due to the large reduction of the reactor thermal power in a short period of time and, thereby, the plant data involve transient uncertainties. The analysis was performed using the RELAP5/MOD1/NSC developed through some modifications of the interphase drag and the wall heat transfer modeling routines of the RELAP5/MOD1/CY018.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.274-284
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• 1992

A 1/15-scale CREARE experiment, which simulates the thermal-hydraulic behavior in the reactor pressure vessel of a PWR during a hypothetical Loss Of Coolant Accident, has been analyzed using CATHARE code for the associated model assessment to represent the phenomenon. The key parameters examined in the CREARE experiment were known as ECC water injection rate. ECC water subcooling, system pressure, and steam flow rate coming out from the core bottom. The present CATHARE simulation, however, has been mainly focused on qualitative analysis of a countercurrent flow in the downcomer. The discrepancy of the simulation results with the experimental data is considered arising primarily from an inadequate numerical representation as well as an interfacial friction model. Accordingly it is suggested from the sensitivity studies that either multidimensional approach or further examination of momentum equations at a junction near a volume element in CATHARE be necessary in order to represent the phenomenon more realistically.

• Proceedings of the Korean Nuclear Society Conference
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• 1995.10a
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• pp.396-402
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• 1995

영광 3,4호기의 부분충수 운전중 정지냉각계통이 상실되고 가압기 Manway가 개방된 사고에 대하여 RELAP5/MOD3.1.2의 열수력 코드를 이용하여 모의하였다. 계산결과 계통의 압력은 최고 1.74bar 까지 도달하였으며, 사고 발생 후 약 1시간 이후부터 계통은 노심이 노출될 때까지 유사 정상상태를 유지한다. 이때 가압기 Manway를 통해 방출되는 증기량은 약 4 kg/s로 붕괴열의 약 80%를 담당하고 증기발생기 2차측에 의해 나머지 20% 가량 제거된다. 또한 비응축성 가스는 계통에 남아 있는 한 계통의 압력 상승율을 증가시키며, RELAP5/MOD3.1.2 계산결과는 일차계통 전체 냉각재의 약 26 %의 질량오차를 나타냈다.