共查询到20条相似文献,搜索用时 31 毫秒
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Antonella L. Costa Patrícia Amélia L. Reis Cláubia Pereira Maria Auxiliadora F. Veloso Amir Z. Mesquita 《Nuclear Engineering and Design》2010,240(6):1487-1494
The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available 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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Patrícia A.L. Reis Antonella L. Costa Claubia Pereira Clarysson A.M. Silva Maria Auxiliadora F. Veloso Amir Z. Mesquita 《Nuclear Engineering and Design》2012
The main aim of this work is to identify how much the code results are affected by the code user in the choice of, for example, the number of thermal hydraulic channels in a nuclear reactor nodalization. To perform this, two essential modifications were made on a previously validated nodalization for analysis of steady-state and forced recirculation off transient in the IPR-R1 TRIGA research reactor. Experimental data were taken as reference to compare the behavior of the reactor for two different types of modeling. The results highlight the necessity of sensitivity analysis to obtain the ideal modeling to simulate a specific system. 相似文献
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Y. Kozmenkov S. Kliem U. Grundmann U. Rohde F.-P. Weiss 《Nuclear Engineering and Design》2007,237(15-17):1938-1951
Plant-measured data provided by the OECD/NEA VVER-1000 coolant transient benchmark programme were used to validate the DYN3D/RELAP5 and DYN3D/ATHLET coupled code systems. Phase 1 of the benchmark (V1000CT-1) refers to an experiment that was conducted during the commissioning of the Kozloduy NPP Unit 6 in Bulgaria. In this experiment, the fourth main coolant pump was switched on whilst the remaining three were running normal operating conditions. The experiment was conducted at 27.5% of the nominal level of the reactor power. 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. The control rods were kept in their original positions during the entire transient. The coupled simulations performed on both DYN3D/RELAP5 and DYN3D/ATHLET were based on the same reactor model, including identical main coolant pump characteristics, boundary conditions, benchmark-specified nuclear data library and nearly identical nodalization schemes. In addition to validation of the coupled code systems against measured data, a code-to-code comparison between simulation results has also been performed to evaluate the respective thermal hydraulic models of the system codes RELAP5 and ATHLET. 相似文献
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An advanced thermal hydraulic code is established on the basis of RELAP5/MOD3.3 code for the investigation of the thermal hydraulic behavior of nuclear power systems. The RELAP5 code is modified by adding a module calculating the effect of rolling motion and introducing new flow and heat transfer models. The experimental data are used to validate the theoretical models and calculation results. It is shown that the advanced flow and heat transfer models could correctly predict the frictional resistance and heat transfer coefficients in rolling motion. The thermal hydraulic code is used to simulate the operation of a natural circulation system in rolling motion. The calculation results are in good agreement with experimental data. The relative discrepancies between calculation results and experimental data are less than 5%. 相似文献
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恰希玛核电厂压力容器各关键部位在PTS瞬态下的温度场研究 总被引:1,自引:1,他引:0
承压热冲击现象在核电厂延寿评估中应被重点关注。本文针对恰希玛核电厂1号机组的压力容器及堆内构件建立了完整的CFD模型,计算了正常工况下压力容器内冷却剂的速度场和温度场分布,计算结果与试验结果符合良好。本文详细研究了蒸汽发生器传热管破裂事故工况下压力容器接管及下降段中冷却剂的热工水力特性,并将计算结果与RELAP5计算结果进行对比,结果表明二者符合良好。本文研究可为反应堆压力容器老化管理评估的计算分析工作提供重要参考。 相似文献
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Pressurized water vessel-type reactor (VVER) safety has become a very important issue, in particular for countries in Central and Eastern Europe. For thermal-hydraulic analyses the western codes like RELAP5, CATHARE and ATHLET were used.The purpose of the study was to quantitatively assess the RELAP5 capability to predict the main circulation pump (MCP) trip at nearly full power transient in Mochovce VVER 440/213 nuclear power plant (NPP). The transient parameters were recorded during the start up test program implementation. For accuracy quantification the improved fast Fourier transform based method (FFTBM) was used. The RELAP5/MOD3.2.2 computer code was used for calculation. The results showed very good agreement between calculated and plant measured data. The results also confirmed some previous studies that the simpler is the transient the higher code accuracy is generally achieved. 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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基于二次开发得到的铅冷快堆一维系统程序RELAP5_LEAD和三维计算流体力学程序FLUENT,利用动态链接库技术和FLUENT用户自定义函数,开发了多尺度耦合分析程序RELAP5/FLUENT。在单相范围内,分别利用耦合程序RELAP5/FLUENT开展简单铅冷串联管道的瞬态流动和传热模拟、简单铅冷闭式回路的瞬态流动模拟,并与RELAP5_LEAD计算结果开展Code-to-Code对比分析。研究结果表明,RELAP5/FLUENT计算结果与RELAP5_LEAD模拟结果吻合良好,耦合程序的开发取得了初步成功,可用于分析铅冷快堆堆内的复杂三维热工水力现象。 相似文献
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The paper presents an evaluation of RELAP5-3D code suitability to model-specific transients that take place during RBMK-1500 reactor operation, where the neutronic response of the core is important. Certain RELAP5-3D transient calculation results were benchmarked against calculation results obtained using the Russian complex neutronic-thermal-hydraulic code STEPAN/KOBRA, specially designed for RBMK reactor analysis. Comparison of the results obtained, using the RELAP5-3D and STEPAN/KOBRA codes, showed reasonable mutual agreement of the calculation results of both codes and their reasonable agreement with the real plant data. 相似文献
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The thermal-hydraulic calculations for the USNRC pressurized thermal shock study, which were performed by the Los Alamos National Laboratory for the Calvert Cliffs nuclear power plant using the TRAC-PF1/MODI code and by the Idaho National Engineering Laboratory for the H.B. Robinson Unit 2 nuclear power plant using the RELAP5/MOD1.6 code, were reviewed at Brookhaven National Laboratory.To quantitatively review these calculations, a simple method based on mass and energy balances was developed at BNL to predict the primary system temperature. In this approach the entire reactor system was lumped into a single volume and the energy balance was applied to that volume. Because significant nonequilibrium effects made it difficult to estimate the pressures, the upper and lower bounds of the pressure were calculated using adiabatic and equilibrium assumptions.In general, the temperatures and pressures of the primary system calculated by both codes were reasonable. The secondary pressures calculated by TRAC indicated it had some difficulty with the condensation model. However, it is not expected that this uncertainty would affect the transient calculations significantly.Review of one typical transient calculation for each plant is discussed in this paper. 相似文献
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This paper deals with the modeling of RBMK-1500 specific transients taking place at Ignalina NPP: measurements of void and fast power reactivity coefficients, as well as change of graphite cooling conditions transient. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and based on the obtained experimental results the actual values of these reactivity coefficients are determined. Graphite temperature reactivity coefficient at the plant is determined by changing graphite cooling conditions in the reactor cavity. This type of transient is unique and important from the point of view of model validation for the gap between fuel channel and the graphite bricks. The measurement results, obtained during this transient, enabled to determine the thermal conductivity coefficient for this gap and to validate the graphite temperature reactivity feedback model. The performed validation of RELAP5-3D model of Ignalina NPP RBMK-1500 reactor allowed to improve the model, which in the future would be used for the safety substantiation calculations of RBMK-1500 reactors. 相似文献
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PWR冷管段1%小破口失水事故实验研究 总被引:1,自引:1,他引:0
在高压综合实验装置(HPITF)上进行核电厂反应堆一次系统冷管段小破口失水事故(SBLOCA)模拟实验,破口方向为冷管段底部,破口面积为1%(NSB-7工况)实验再现了核电厂发生小破口失水事故时的热工水力学现象,实验结果与RELAP5/MOD2分析程序的计算结果上比较,验证了该程序对小破口失水事故的分析能力。 相似文献