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本文目的是应用最佳估算(BE)程序来确定某型研究堆的一个重要的安全裕量,即流动不稳定起始点(OFI)。这些BE程序系统针对动力堆开发,并且利用该类反应堆运行条件下的相关实验数据进行了校核,但它们不能直接应用于研究堆。近几年,对这些BE程序作了改进,并建议应用于研究堆。因此,本文的工作是要研究RELAP5/Modd程序系统预测在研究堆运行条件下(低温、低压)均匀加热垂直燃料板的流动不稳定起始点(0H)的能力。为此,选择了橡树岭国家实验室的热工水力试验回路(ORNL-THTL)进行改造,这个热工水力回路是为了反映ORNL的先进中子源反应堆(ANSR)的运行条件而设计建造的。
在这一工作框架下获得的结果显示:RELAP5/Mod3.2在预测OFI的临界区域时存在一定的局限,并且计算所得数据没能验证质量流速和热流密度的封闭关系式。 相似文献
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低干度自然循环流量漂移的特征曲线图谱分析 总被引:1,自引:0,他引:1
在5MW低温核供热堆全模拟试验回路(HRTL-5)上,实验观察到了低干度自然循环条件下的流量漂移现象.通过一个考虑了加热段欠热沸腾、上升段冷凝、闪蒸等物理过程的两相流动数学模型,编制了相应的计算程序,获得了自然循环特征曲线图谱及其运行曲线,确定了自然循环分岔图和静态不稳定边界图,进而提出了通过自然循环特征曲线图谱研究流量漂移的分析方法.分析表明:特征曲线图谱方法是研究自然循环静态不稳定的有效手段.增大系统压力、减小热流密度、增加入口单相阻力、减小出口两相阻力有利于避免自然循环流量漂移的发生. 相似文献
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喷泉不稳定诱发间歇流量振荡实验研究 总被引:5,自引:3,他引:2
实验在5MW低温核供热堆热工水力学模拟回路上完成,系统压力为0.1MPa。研究了发生间歇流量振荡的条件及机理。给出了喷泉不稳定诱发的间歇流量振荡的物理模型。 相似文献
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本文根据5MW低温供热堆的实验和运行经验,分析了该供热系统的特点,并据此提出了对这种类型的供热系统负荷跟踪及核功率自动控制的方案,以调节负载为主,调节反应堆功率为辅;即控制二回路流量变化来调节热网温度,调节核功率仅用以维持二回路的温度。 相似文献
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利用STEADY-LHTR程序,对清华大学核能技术设计研究院所设计的200MW核供热堆的两相流动系统的稳定性、并联通道流动不稳定性的现象作了描述和机理分析。对200MW核供热堆自然循环系统流动特性作了大量的分析计算,计算结果以表图形式给出。计算结果表明,①200MW核供热堆自然循环的流量随堆芯入口温度的升高而稍有增加。②额定设计工况下,反应堆的自然循环系统有很好的流动稳定性。③在额定压力2.0MPa下,堆芯入口温度接近155℃时,自然循环系统可能出现莱迪内格不稳定及平行通道不稳定流动。 相似文献
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《Annals of Nuclear Energy》2005,32(4):379-397
In this paper, two-phase flow instability in natural circulation loops of China Advanced Research Reactor (CARR) has been investigated. CARR is a low pressure and low power density research reactor. A natural circulation instability analysis model is developed for the natural circulation loop of CARR. The homogeneous flow model is used to establish the system control equations. The non-uniform heating and subcooled boiling heat transfer is included. The accumulation heat of the wall is also included. Numerical method of Gear is employed to solve the system equations documented in terms of ordinary differential equations. According to the calculation results, stability maps of the natural circulation loop, which confirm the presence of an instability region under the conditions of low equilibrium quality in the outlet and low pressure, are obtained. It is a special kind of density wave oscillation (DWO) that occurs in very low equilibrium quality region with the characteristics of geysering and ‘Type-I’ DWO at the same time. The calculation results show such oscillation course clearly. The variations of the mass flow rate, the pressure drop and the boiling boundary are analyzed separately. Especially, the phase-space trajectory of the boiling boundary and the mass flow rate is discussed. Finally the oscillation frequency is discussed. The calculated results have important significance for the safety operation and accidental analysis of CARR. 相似文献
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Tewfik Hamidouche Nouara Rassoul El-Khider Si-Ahmed Hakim El Hadjen Anis Bousbia-salah 《Progress in Nuclear Energy》2009,51(3):485-495
Establishment of safety margins and the corresponding operating condition limits will ensure achievement of a safe operation of nuclear installations. For this purpose, several critical phenomena have been analyzed theoretically and experimentally and a great number of models and correlations are made available. Among these critical issues the well-known flow instability has been intensively investigated by several authors especially for nuclear power plants' (NPPs) operating conditions. However, limited published work is available for research reactor operation conditions. In general, the Whittle and Forgan correlation is widely used to define the margin to static flow instabilities in narrow parallel heated channels for research reactors.In the framework of verification and assessment of the capabilities of the RELAP5/Mod 3 system code to determine the onset of flow instability in research reactor conditions, a simple model based on steady-state equations adjusted with drift-flux correlations has been developed. The program is used to draw the pressure drop characteristic curves and to establish the conditions of the Ledinegg instability in a uniformly heated channel subject to constant outlet pressure. The model is assessed by using experimental data from a thermal hydraulic test loop by Siman-Tov and numerical results from RELAP5/Mod 3. The model presents acceptable estimation of the target mass flow that would induce flow instability and the latter could be then used to establish a conservative margin to the Ledinegg instability. 相似文献
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蒸汽发生器传热管内伴有两相流的产生,研究两相情况下传热管内密度波不稳定现象,对控制核反应堆安全运行有着至关重要的作用。通过数值计算,研究了双侧对流换热条件下的传热管密度波震荡(DWO)现象。引用Babcock&Wilcox公司的直流式蒸汽发生器(OTSG)实验进行计算模型的可靠性验证;将传热管双侧对流换热与壁面均匀加热条件下的流动不稳定现象进行比较;分析管内流体加热段高度、流动方向变化时,不稳定边界的移动情况。结果表明,增加加热段高度、适当减少传热管水平方向倾斜角度(50°~90°内变化)可以增加系统的稳定性。该研究可以为螺旋管式蒸汽发生器设计提供参考。 相似文献
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For the problem of two-phase natural circulation flow in gap clearance between reactor vessel lower head and insulator in the condition of severe accident, one-dimensional steady-state natural flow analysis code was written by utilizing FORTRAN. Based on the code, the effects of different correlations for friction coefficient and the number of nodes of heating section on mass flow rate of two-phase natural circulation flow were studied. And the results are compared with that of Chinese REPEC experiment and simulation using RELAP5 program so as to verify the rationality and correctness of the code. Based on the experiment data, simulation results and the model, friction coefficient and the void fraction condition under ERVC correlation are obtained by fitting. The results calculated by the model using fitting friction coefficient correlation agree well with ULPU V test data. Furthermore, the effect of power, pressure, inlet area, gap diameter, flooding level and inlet water subcooling on mass flow rate and void fraction of two-phase natural circulation were studied utilizing this code. 相似文献
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《Annals of Nuclear Energy》2007,34(1-2):1-12
Correct prediction of water hammer transients is of paramount importance for the safe operation of the plant. Therefore, verification of computer codes capability to simulate water hammer type transients is a very important issue at performing safety analyses for nuclear power plants. Verification of RELAP5/MOD3.3 code capability to simulate water hammer type transients employing the experimental investigations is presented. Experience gained from benchmarking analyses has been used at development of the detail RELAP5 code RBMK-1500 model for simulation of water hammer effects in reactor main circulation circuit. In RBMK-type reactors the water hammers can occur in cases of rapid check valve operation. The performed analysis using RELAP5 code RBMK-1500 model has shown that in general the maximum values of the pressure pulses due to water hammer do not exceed the permissible loads on the pipelines. 相似文献
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Nowadays, the coupled codes technique, which consists in incorporating three-dimensional (3D) neutron modeling of the reactor core into system codes, is extensively used for carrying out best estimate (BE) simulation of complex transient in nuclear power plants (NPP). This technique is particularly suitable for transients that involve core spatial asymmetric phenomena and strong feedback effects between core neutronics and reactor loop thermal-hydraulics. Such complex interactions are encountered under normal and abnormal operating conditions of a boiling water reactors (BWR). In such reactors Oscillations may take place owing to the dynamic behavior of the liquid-steam mixture used for removing the thermal power. Therefore, it is necessary to be able to detect in a reliable way these oscillations. The purpose of this work is to characterize one aspect of these unstable behaviors using the coupled codes technique. The evaluation is performed against Peach Bottom-2 low-flow stability tests number 3 using the coupled RELAP5/PARCS code. In this transient dynamically complex neutron kinetics coupling with thermal-hydraulics events take place in response to a core pressure perturbation. The calculated coupled code results are herein assessed and compared against the available experimental data. 相似文献
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Plant-measured data provided within the specification of the OECD/NEA VVER-1000 coolant transient benchmark (V1000CT) were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to the MCP (main coolant pump) switching on experiment conducted in the frame of the plant-commissioning activities at the Kozloduy NPP Unit 6 in Bulgaria. The experiment was started at the beginning of cycle (BOC) with average core expose of 30.7 effective full power days (EFPD), when the reactor power was at 27.5% of the nominal level and three out of four MCPs were operating. The transient is characterized by a rapid increase in the primary coolant flow through the core and, as a consequence, a decrease of the space-dependent core inlet temperature. Both DYN3D/RELAP5 and DYN3D/ATHLET analyses were based on the same reactor model, including identical MCP characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. For an adequate modelling of the redistribution of the coolant flow in the reactor pressure vessel during the transient a simplified mixing model for the DYN3D/ATHLET code was developed and validated against a computational fluid dynamics calculation.
The results of both coupled code calculations are in good agreement with the available experimental data. The discrepancies between experimental data and the results of both coupled code calculations do not exceed the accuracy of the measurement data. This concerns the initial steady-state data as well as the time histories during the transient. In addition to the validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has been performed to evaluate relevant thermal hydraulic models of the system codes RELAP5 and ATHLET and to explain differences between the calculation results. 相似文献
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