• 제목/요약/키워드: Natural circulation system

검색결과 218건 처리시간 0.033초

Prediction of dryout-type CHF for rod bundle in natural circulation loop under motion condition

  • Huang, Siyang;Tian, Wenxi;Wang, Xiaoyang;Chen, Ronghua;Yue, Nina;Xi, Mengmeng;Su, G.H.;Qiu, Suizheng
    • Nuclear Engineering and Technology
    • /
    • 제52권4호
    • /
    • pp.721-733
    • /
    • 2020
  • In nuclear engineering, the occurrence of critical heat flux (CHF) is complicated for rod bundle, and it is much more difficult to predict the CHF when it is in natural circulation under motion condition. In this paper, the dryout-type CHF is investigated for the rod bundle in a natural circulation loop under rolling motion condition based on the coupled analysis of subchannel method, a one-dimensional system analysis method and a CHF mechanism model, namely the three-fluid model for annular flow. In order to consider the rolling effect of the natural circulation loop, the subchannel model is connected to the one-dimensional system code at the inlet and outlet of the rod bundle. The subchannel analysis provides the local thermal hydraulic parameters as input for the CHF mechanism model to calculate the occurrence of CHF. The rolling motion is modeled by additional motion forces in the momentum equation. First, the calculation methods of the natural circulation and CHF are validated by a published natural circulation experiment data and a CHF empirical correlation, respectively. Then, the CHF of the rod bundle in a natural circulation loop under both the stationary and rolling motion condition is predicted and analyzed. According to the calculation results, CHF under stationary condition is smaller than that under rolling motion condition. Besides, the CHF decreases with the increase of the rolling period and angular acceleration amplitude within the range of inlet subcooling and mass flux adopted in the current research. This paper can provide useful information for the prediction of CHF in natural circulation under motion condition, which is important for the nuclear reactor design improvement and safety analysis.

CUPID 코드의 유체 물성치 변화를 고려한 자연대류 해석 (NATURAL CIRCULATION ANALYSIS CONSIDERING VARIABLE FLUID PROPERTIES WITH THE CUPID CODE)

  • 이승준;박익규;윤한영;김정우
    • 한국전산유체공학회지
    • /
    • 제20권4호
    • /
    • pp.14-20
    • /
    • 2015
  • Without electirc power to cool down the hot reactor core, passive systems utilizing natural circulation are becoming a big specialty of recent neculear systems after the severe accident in Fukusima. When we consider the natural circulation in a pool, thermal mixing phenomena may start from single phase circulation and can continue to two phase condition. Since the CUPID code, which has been developed for two-phase flow analysis, can deal with the phase transition phenomena, the CUPID would be pertinent to natural convection problems in single- and two-phase conditions. Thus, the CUPID should be validated against single- and two-phase natural circulation phenomena. For the first step of the validation process, this study is focused on the validation of single-phase natural circulation. Moreover, the CUPID code solves the fluid properties by the relationship to pressure and temperature from the steam table considering non-condensable gas effects, so that the effects from variable properties are included. Simple square thermal cavity problems are tested for laminar and turbulent conditions against numerical and experimental data. Throughout the investigation, it is found that the variable properties can affect the flow field in laminar condition, but the effect becomes weak in turbulence condition, and the CUPID code implementing steam table is capable of analyzing single phase natural circualtion phenomena.

원자로용기 외벽냉각시 용기와 단열재 사이의 자연순환 이상유동에 관한 실험적 연구 (An Experimental Study on the Two-Phase Natural Circulation Flow through an Annular Gap between Reactor Vessel and Insulation under External Vessel Cooling)

  • 하광순;박래준;김환열;김상백;김희동
    • 대한기계학회:학술대회논문집
    • /
    • 대한기계학회 2003년도 춘계학술대회
    • /
    • pp.1897-1902
    • /
    • 2003
  • An 1/21.6 scaled experimental facility was prepared utilizing the results of a scaling analysis to simulate the APRI400 reactor and insulation system. The behaviors of the boiling-induced two-phase natural circulation flow in the insulation gap were observed, and the liquid mass flow rates driven by natural circulation loop were measured by varying the wall heat flux, upper exit slot area and configuration. And non-heating experiments have also been performed and discussed to certify the hydraulic similarity of the heating experiments by injecting air equivalent to the steam generated in the heating experimental condition.

  • PDF

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

  • Zarei, M.
    • Nuclear Engineering and Technology
    • /
    • 제51권5호
    • /
    • pp.1272-1278
    • /
    • 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.

소형 자연순환 수직형 증발기 해석 및 성능실험 (Performance Analysis and Test of a Small-Scale Natural Circulation Vertical Evaporator)

  • 차상진;김내현;유진상
    • 설비공학논문집
    • /
    • 제23권9호
    • /
    • pp.595-603
    • /
    • 2011
  • In this study, an effort has been made to analyze the subcooled boiling heat transfer in a natural circulation vertical evaporator. To verify the analysis, a small-scale model was made and tested. The friction correlation by Ueda, void fraction and quality correlation by Saha and Zuber along with the superposition heat transfer model by Rohsenow yielded a satisfactory agreement with the model test data. The analysis was extended to simulate a 1 ton/day concentration system. Comparison with the test results of 1 ton/day prototype revealed that the data were overpredicted by 13%. The capacity of the prototype was 1.2 ton/day with COP of 5.77.

Experimental and numerical investigations on effect of reverse flow on transient from forced circulation to natural circulation

  • Li, Mingrui;Chen, Wenzhen;Hao, Jianli;Li, Weitong
    • Nuclear Engineering and Technology
    • /
    • 제52권9호
    • /
    • pp.1955-1962
    • /
    • 2020
  • In a sudden shutdown of primary pump or coolant loss accident in a marine nuclear power plant, the primary flow decreases rapidly in a transition process from forced circulation (FC) to natural circulation (NC), and the lower flow enters the steam generator (SG) causing reverse flow in the U-tube. This can significantly compromise the safety of nuclear power plants. Based on the marine natural circulation steam generator (NCSG), an experimental loop is constructed to study the characteristics of reverse flow under middle-temperature and middle-pressure conditions. The transition from FC to NC is simulated experimentally, and the characteristics of SG reverse flow are studied. On this basis, the experimental loop is numerically modeled using RELAP5/MOD3.3 code for system analysis, and the accuracy of the model is verified according to the experimental data. The influence of the flow variation rate on the reverse flow phenomenon and flow distribution is investigated. The experimental and numerical results show that in comparison with the case of adjusting the mass flow discontinuously, the number of reverse flow tubes increases significantly during the transition from FC to NC, and the reverse flow has a more severe impact on the operating characteristics of the SG. With the increase of flow variation rate, the reverse flow is less likely to occur. The mass flow in the reverse flow U-tubes increases at first and then decreases. When the system is approximately stable, the reverse flow is slightly lower than obverse flow in the same U-tube, while the flow in the obverse flow U-tube increases.

Conceptual design of small modular reactor driven by natural circulation and study of design characteristics using CFD & RELAP5 code

  • Kim, Mun Soo;Jeong, Yong Hoon
    • Nuclear Engineering and Technology
    • /
    • 제52권12호
    • /
    • pp.2743-2759
    • /
    • 2020
  • A detailed computational fluid dynamics (CFD) simulation analysis model was developed using ANSYS CFX 16.1 and analyzed to simulate the basic design and internal flow characteristics of a 180 MW small modular reactor (SMR) with a natural circulation flow system. To analyze the natural circulation phenomena without a pump for the initial flow generation inside the reactor, the flow characteristics were evaluated for each output assuming various initial powers relative to the critical condition. The eddy phenomenon and the flow imbalance phenomenon at each output were confirmed, and a flow leveling structure under the core was proposed for an optimization of the internal natural circulation flow. In the steady-state analysis, the temperature distribution and heat transfer speed at each position considering an increase in the output power of the core were calculated, and the conceptual design of the SMR had a sufficient thermal margin (31.4 K). A transient model with the output ranging from 0% to 100% was analyzed, and the obtained values were close to the Thot and Tcold temperature difference value estimated in the conceptual design of the SMR. The K-factor was calculated from the flow analysis data of the CFX model and applied to an analysis model in RELAP5/MOD3.3, the optimal analysis system code for nuclear power plants. The CFX analysis results and RELAP analysis results were evaluated in terms of the internal flow characteristics per core output. The two codes, which model the same nuclear power plant, have different flow analysis schemes but can be used complementarily. In particular, it will be useful to carry out detailed studies of the timing of the steam generator intervention when an SMR is activated. The thermal and hydraulic characteristics of the models that applied porous media to the core & steam generators and the models that embodied the entire detail shape were compared and analyzed. Although there were differences in the ability to analyze detailed flow characteristics at some low powers, it was confirmed that there was no significant difference in the thermal hydraulic characteristics' analysis of the SMR system's conceptual design.

Unsteady Single-Phase Natural Circulation Flow Mixing Prediction Using CATHARE Three-Dimensional Capabilities

  • Salah, Anis Bousbia;Vlassenbroeck, Jacques
    • Nuclear Engineering and Technology
    • /
    • 제49권3호
    • /
    • pp.466-475
    • /
    • 2017
  • Coolant mixing under natural circulation flow regime constitutes a key parameter that may play a role in the course of an accidental transient in a nuclear pressurized water reactor. This issue has motivated some experimental investigations carried out within the OECD/NEA PKL projects. The aim was to assess the coolant mixing phenomenon in the reactor pressure vessel downcomer and the core lower plenum under several asymmetric steady and unsteady flow conditions, and to provide experimental data for code validations. Former studies addressed the mixing phenomenon using, on the one hand, one-dimensional computational approaches with cross flows that are not fully validated under transient conditions and, on the other hand, expensive computational fluid dynamic tools that are not always justified for large-scale macroscopic phenomena. In the current framework, an unsteady coolant mixing experiment carried out in the Rossendorf coolant mixing test facility is simulated using the three-dimensional porous media capabilities of the thermal-hydraulic system CATHARE code. The current study allows highlighting the current capabilities of these codes and their suitability for reproducing the main phenomena occurring during asymmetric transient natural circulation mixing conditions.

Investigation of Characteristics of Passive Heat Removal System Based on the Assembled Heat Transfer Tube

  • Wu, Xiangcheng;Yan, Changqi;Meng, Zhaoming;Chen, Kailun;Song, Shaochuang;Yang, Zonghao;Yu, Jie
    • Nuclear Engineering and Technology
    • /
    • 제48권6호
    • /
    • pp.1321-1329
    • /
    • 2016
  • To get an insight into the operating characteristics of the passive residual heat removal system of molten salt reactors, a two-phase natural circulation test facility was constructed. The system consists of a boiling loop absorbing the heat from the drain tank, a condensing loop consuming the heat, and a steam drum. A steady-state experiment was carried out, in which the thimble temperature ranged from $450^{\circ}C$ to $700^{\circ}C$ and the system pressure was controlled at levels below 150 kPa. When reaching a steady state, the system was operated under saturated conditions. Some important parameters, including heat power, system resistance, and water level in the steam drum and water tank were investigated. The experimental results showed that the natural circulation system is feasible in removing the decay heat, even though some fluctuations may occur in the operation. The uneven temperature distribution in the water tank may be inevitable because convection occurs on the outside of the condensing tube besides boiling with decreasing the decay power. The instabilities in the natural circulation loop are sensitive to heat flux and system resistance rather than the water level in the steam drum and water tank. RELAP5 code shows reasonable results compared with experimental data.

자연순환식 태양열 급탕 시스템의 성능 추정 방법에 관한 연구 (A Study on the System Performance Prediction Method of Natural Circulation Solar Hot Water System)

  • 윤석범;전문헌
    • 태양에너지
    • /
    • 제7권2호
    • /
    • pp.37-53
    • /
    • 1987
  • This study has been prepared for the purpose of developing the system performance prediction method of natural circulation solar hot water system. The storage tank of the natural circulation solar hot water system equipped with flat-plate solar collector is located at higher elevation than the solar collectors. Therefor, the storage tank temperature distribution formed accordance with configuration of storage tank by flow rate of circulating fluid affect system collection efficiency. In this study measure the storage tank temperature distribution with various experimental system under real sun condition and present the theoretical prediction method of the storage tank temperature. Moreover measure the flow rate not only day-time but also night-time reverse flow rate with die injection visual flow meter. Main conclusion obtain from the present study is as follows; 1) The storage tank temperature distribution above the connecting pipe connection position is the same as that of the fully mixed tank and below the connection position is the same as that of stratified tank. 2) The system performance sensitive to the storage tank temperature distribution. Therefore detailed tank model is necessary. Average storage tank temperature can be calculate 3% and storage tank temperature profile can get less than 10% difference with this model system.

  • PDF