• 제목/요약/키워드: Coolant Flow Analysis

검색결과 259건 처리시간 0.021초

공기와 물의 이상 자연순환 유동의 1 차원 해석 (One-Dimensional Analysis of Air-Water Two Phase Natural Circulation Flow)

  • 박래준;하광순;김재철;홍성완;김상백
    • 대한기계학회:학술대회논문집
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    • 대한기계학회 2007년도 춘계학술대회B
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    • pp.2626-2631
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    • 2007
  • Air-water two phase natural circulation flow in the T-HERMES (Thermo-Hydraulic Evaluation of Reactor cooling Mechanism by External Self-induced flow)-1D experiment has been evaluated to verify and evaluate the experimental results by using the RELAP5/MOD3 computer code. The RELAP5 results have shown that an increase in the coolant inlet area leads to an increase in the water circulation mass flow rate. However, the water outlet area does not effective on the water circulation mass flow rate. As the coolant outlet moves to a lower position, the water circulation mass flow rate decreases. The water level is not effective on the water circulation mass flow rate. As the height increases in the air injection part, the void fraction increases. However, the void fraction in the upper part of the air injector maintains a constant value. An increase in the air injection mass flow rate leads to an increase in the local void fraction, but it is not effective on the local pressure.

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전동 워터펌프의 열유동 특성 해석에 관한 연구 (A Study on Thermo-flow Characteristics Analysis of Electric Water Pump)

  • 김성철;송형근
    • 한국자동차공학회논문집
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    • 제20권5호
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    • pp.95-101
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    • 2012
  • An electric water pump for engine cooling system has an advantage which particularly in the cold start, the use of the electric water pump saves fuel and leads to a corresponding reduction in emissions. The canned type electric water pump without mechanical sealing elements was selected to meet the requirements for operational reliability and life. However, the electric water pump for internal combustion engine generates much more heat loss than for hybrid electric vehicle since it is operated by the electric power of high current and low voltage. In this study, the fluid flow and thermal characteristics of the canned type electric water pump as an inverter integrated water pump has been investigated under the effects of heat generation. The analysis conditions such as outdoor air temperature of $125^{\circ}C$, water pump speed of 6000 rpm, coolant temperature of $106^{\circ}C$ and coolant flow rate of 120 L/min was used as a standard condition. Therefore, flow fields and temperature distribution inside the water pump were obtained. Also, we checked the feasibility of the canned type for the electric water pump in comparison with the mechanical seal type.

Flow Characteristics Analysis for the Chemical Decontamination of the Kori-1 Nuclear Power Plant

  • Cho, Seo-Yeon;Kim, ByongSup;Bang, Youngsuk;Kim, KeonYeop
    • 방사성폐기물학회지
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    • 제19권1호
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    • pp.51-58
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    • 2021
  • Chemical decontamination of primary systems in a nuclear power plant (NPP) prior to commencing the main decommissioning activities is required to reduce radiation exposure during its process. The entire process is repeated until the desired decontamination factor is obtained. To achieve improved decontamination factors over a shorter time with fewer cycles, the appropriate flow characteristics are required. In addition, to prepare an operating procedure that is adaptable to various conditions and situations, the transient analysis results would be required for operator action and system impact assessment. In this study, the flow characteristics in the steady-state and transient conditions for the chemical decontamination operations of the Kori-1 NPP were analyzed and compared via the MARS-KS code simulation. Loss of residual heat removal (RHR) and steam generator tube rupture (SGTR) simulations were conducted for the postulated abnormal events. Loss of RHR results showed the reactor coolant system (RCS) temperature increase, which can damage the reactor coolant pump (RCP)s by its cavitation. The SGTR results indicated a void formation in the RCS interior by the decrease in pressurizer (PZR) pressure, which can cause surface exposure and tripping of the RCPs unless proper actions are taken before the required pressure limit is achieved.

막냉각이 적용된 액체로켓엔진의 비추력 최적조건 (Optimal Condition of Specific Impulse for a Liquid Rocket Engine with Film Cooling)

  • 조원국;박순영;설우석
    • 한국추진공학회:학술대회논문집
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    • 한국추진공학회 2007년도 제28회 춘계학술대회논문집
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    • pp.135-140
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    • 2007
  • 막냉각이 적용된 액체로켓엔진의 비추력 최적조건을 분석하였다. 본 연구에서 구한 엔진의 비추력 성능을 문헌에 소개된 가스발생기 사이클 엔진의 개념설계와 비교하여 적절한 정확성을 가지는 것을 확인하였다. 비추력을 극대화할 수 있는 조건으로 최적의 막냉각 유량과 재생냉각 용량의 조합이 제시되었다. 추력이 증가될 경우, 막냉각 유량이 감소하고 연료펌프 차압은 증가한다. 터빈 입구온도 증가에 따라 최적 조건의 막냉각 유량과 연료펌프 차압이 증가한다. 코킹 온도 증가 역시 터빈 입구온도와 정성적으로 동일한 영향을 가진다.

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원형 T분기배관 내 누설유동의 열성층화와 난류침투에 관한 전산해석적 연구 (Numerical Analysis of Thermal Stratification and Turbulence Penetration into Leaking Flow in a Circular Branch Piping)

  • 한성민;최영돈
    • 대한기계학회:학술대회논문집
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    • 대한기계학회 2003년도 춘계학술대회
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    • pp.1833-1838
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    • 2003
  • In the nuclear power plant, emergency core coolant system(ECCS) is furnished at reactor coolant system(RCS) in order to cool down high temperature water in case of emergency. However, in this coolant system, thermal stratification phenomenon can be occurred due to coolant leaking in the check valve. The thermal stratification produces excessive thermal stresses at the pipe wall so as to yield thermal fatigue crack(TFC) accident. In the present study, when the turbulence penetration occurs in the branch piping, the maximum temperature differences of fluid at the pipe cross-sections of the T-branch with thermal stratification are examine

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ROSA/LSTF test and RELAP5 code analyses on PWR 1% vessel upper head small-break LOCA with accident management measure based on core exit temperature

  • Takeda, Takeshi
    • 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.

핵연료다발 유동혼합 날개 개발을 위한 CFD 응용 (CFD Application to Development of Flow Mixing Vane in a Nuclear Fuel Assembly)

  • 인왕기;오동석;전태현
    • 대한기계학회:학술대회논문집
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    • 대한기계학회 2001년도 춘계학술대회논문집E
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    • pp.482-487
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    • 2001
  • A CFD study was conducted to evaluate the nuclear fuel assembly coolant mixing that is promoted by the flow-mixing vanes on the grid spacer. Four mixing vanes (split vane, swirl vane, twisted vane, hybrid vane) were chosen in this study. A single subchannel of one grid span is modeled using the flow symmetry. The three mixing vanes other than swirl vane generate a large crossflow between the subchannels and a skewed elliptic swirling flow in the subchannel near the grid spacer. The swirl vane induces a circular swirling flow in the subchannel and a negligible crossflow. The split vane and the twisted vane were predicted to result in relatively larger pressure drop across the grid spacer. Since the average turbulent kinetic energy in the subchannel rapidly decreases to a fully developed level downstream of the spacer, turbulent mixing caused by the mixing vanes appears to be not as effective as swirling flow mixing in the subchannel. In summary, the CFD analysis represented the overall characteristics of coolant mixing well in a nuclear fuel assembly with the flow mixing vanes on the grid spacer. The CFD study is therefore quite useful for the development of an advanced flow-mixing vane.

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Research on the structure design of the LBE reactor coolant pump in the lead base heap

  • Lu, Yonggang;Zhu, Rongsheng;Fu, Qiang;Wang, Xiuli;An, Ce;Chen, Jing
    • Nuclear Engineering and Technology
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    • 제51권2호
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    • pp.546-555
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    • 2019
  • Since the first nuclear reactor first critical, nuclear systems has gone through four generations of history, and the fourth generation nuclear system will be truly realized in the near future. The notions of SVBR and lead-bismuth eutectic alloy coolant put forward by Russia were well received by the international nuclear science community. Lead-bismuth eutectic alloy with the ability of the better neutron economy, the low melting point, the high boiling point, the chemical inertness to water and air and other features, which was considered the most promising coolant for the 4th generation nuclear reactors. This study mainly focuses on the structural design optimization of the 4th-generation reactor coolant pump, including analysis of external characteristics, inner flow, and transient characteristic. It was found that: the reactor coolant pump with a central symmetrical dual-outlet volute structure has better radial-direction balance, the pump without guide vane has better hydraulic performance, and the pump with guide vanes has worse torsional vibration and pressure pulsation. This study serves as experience accumulation and technical support for the development of the 4th generation nuclear energy system.

An approach to the coupled dynamics of small lead cooled fast reactors

  • Zarei, M.
    • Nuclear Engineering and Technology
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    • 제51권5호
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    • pp.1272-1278
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    • 2019
  • A lumped kinetic modeling platform is developed to investigate the coupled nuclear/thermo-fluid features of the closed natural circulation loop in a low power lead cooled fast reactor. This coolant material serves a reliable choice with noticeable thermo-physical safety characteristics in terms of natural convection. Boussienesq approximation is resorted to appropriately reduce the governing partial differential equations (PDEs) for the fluid flow into a set of ordinary differential equations (ODEs). As a main contributing step, the coolant circulation speed is accordingly correlated to the loop operational power and temperature levels. Further temporal analysis and control synthesis activities may thus be carried out within a more consistent state space framework. Nyquist stability criterion is thereafter employed to carry out a sensitivity analysis for the system stability at various power and heat sink temperature levels and results confirm a widely stable natural circulation loop.

Thermal-Mixing Analyses for Safety Injection at Partial Loop Stagnation of a Nuclear Power Plant

  • Hwang, Kyung-Mo;Kim, Kyung-Hoon
    • Journal of Mechanical Science and Technology
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    • 제17권9호
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    • pp.1380-1387
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    • 2003
  • When a cold HPSI (High pressure Safety Injection) fluid associated with an overcooling transient, such as SGTR (Steam Generator Tube Rupture), MSLB (Main Steam Line Break) etc., enters the cold legs of a stagnated primary coolant loop, thermal stratification phenomena will arise due to incomplete mixing. If the stratified flow enters the downcomer of the reactor pressure vessel, severe thermal stresses are created in a radiation embrittled vessel wall by local overcooling. As general thermal-hydraulic system analysis codes cannot properly predict the thermal stratification phenomena, RG 1.154 requires that a detailed thermal-mixing analysis of PTS (pressurized Thermal Shock) evaluation be performed. Also. previous PTS studies have assumed that the thermal stratification phenomena generated in the stagnated loop side of a partially stagnated primary coolant loop are neutralized in the vessel downcomer by the strong flow from the unstagnated loop. On the basis of these reasons, this paper focuses on the development of a 3-dimensional thermal-mixing analysis model using PHOENICS code which can be applied to both partial and total loop stagnated cases. In addition, this paper verifies the fact that, for partial loop stagnated cases, the cold plume generated in the vessel downcomer due to the thermal stratification phenomena of the stagnated loop is almost neutralized by the strong flow of the unstagnated loop but is not fully eliminated.