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中欧核能合作研究项目超临界水堆燃料验证实验(SCWR-FQT)的主要研究内容为在超临界水环境下对一个小型燃料组件进行堆内性能分析和验证。本文应用修过后的系统程序ATHLET-SC对该实验回路进行建模,同时结合堆芯中子物理的计算结果,对由于压力管进口管破裂形成的失水事故进行热工水力和中子物理的耦合分析,并讨论了物理耦合中停堆棒的负反应性、冷却剂温度系数等参数对结果的影响。计算结果表明,进行了中子物理耦合的结果得到的最高包壳温度比未进行中子耦合的结果要低15℃,同时停堆棒引入的负反应性是该事故过程中影响燃料棒最高包壳温度的一个主要因素。 相似文献
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针对超临界水冷反应堆(SCWR)堆芯冷却剂密度沿轴向变化剧烈的特点,开发用于SCWR堆芯稳态物理-热工水力耦合计算的程序系统CASIR。CASIR由改进的压水堆堆芯中子学计算程序和适用于SCWR燃料组件计算的子通道热工-水力程序组成,具备调整堆芯下腔室入口流量分配的功能。针对CSR1000双流程的SCWR首循环堆芯,通过与蒙特卡罗程序对比寿期初时刻计算结果的方式,初步验证CASIR计算SCWR堆芯中子学问题的准确性;通过SCWR堆芯燃耗模拟,以及调整堆芯流量分布使得最大包壳表面温度(MCST)满足设计限值的测试,表明CASIR满足SCWR堆芯设计的要求,可应用于方形燃料组件的SCWR堆芯概念设计。 相似文献
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《核动力工程》2015,(4):41-44
基于节块法中子扩散计算程序,二次开发了具备调棒临界-燃耗计算及燃料管理能力的超临界水堆(SCWR)堆芯稳态中子学计算程序NGFMN_S。通过模块化方式耦合NGFMN_S和超临界水堆子通道热工-水力计算程序ATHAS,开发了超临界水堆堆芯三维物理-热工水力耦合稳态性能分析程序SNTA。针对超临界水堆堆芯CSR1000,通过与耦合程序CASIR及SRAC/SPROD对比检验,结果表明:SNTA程序针对CSR1000问题的计算结果与参考程序符合良好;相比于堆芯计算采用细网有限差分方法的CASIR或SRAC/SPROD程序,SNTA程序的计算效率显著提高;适用于具备强烈核热耦合特性的超临界水堆堆芯的稳态性能分析。 相似文献
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开发了THAS-PC2子通道分析微机程序,用于计算稳态和瞬态工况下快堆燃料组件的流量、温度和压力等参量分布。对EFR燃料组件的稳态和瞬态计算结果如下:堆芯出口平均温度和温长分别为557℃和157℃,最高包壳表面温度为601℃,它发生在中心燃料棒上,最大冷却剂温度为593℃;主泵断电二次停堆事故作为瞬态计算,算得的最高冷却剂温度和最高包壳表面温度分别为630℃和637℃(当t=2s时),它们都远低于 相似文献
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针对超临界水堆(SCWR)控制棒落入堆芯事件特点,采用堆芯三维瞬态性能分析方法,利用开发的SCWR堆芯三维瞬态物理-热工水力耦合程序STTA,建立SCWR堆芯落棒瞬态三维计算模型和分析流程,研究分析超临界水堆CSR1000在控制棒落入堆芯瞬态过程中的堆芯性能,分析评价落棒瞬态下CSR1000堆芯的安全性能。堆芯三维落棒瞬态分析表明,当落入堆芯棒束价值较高时,落棒初期堆芯功率下降较快,之后由于水密度的反应性反馈,堆芯功率缓慢回升至新的平衡,堆芯功率下降速率超过了停堆信号整定值,将触发保护停堆;当落入堆芯棒束价值较低时,由于水密度的反应性反馈,堆芯功率下降缓慢,堆芯功率下降速率未能达到停堆信号整定值,不能触发保护停堆。控制棒落入堆芯对堆芯轴向功率分布影响很小,高价值落棒导致的落棒区域燃料组件功率坍塌相对低价值落棒更明显。无论是高价值落棒还是低价值落棒,瞬态过程中最大包壳壁面温度均低于瞬态安全限值850℃。水密度的显著反应性反馈及必要的保护停堆措施能保证CSR1000堆芯在控制棒落入堆芯过程中的安全性能。 相似文献
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本文建立了中国先进研究堆标准燃料组件单组件的流-固耦合共轭传热CFD分析模型。通过1组稳态流量工况的分析,拟合获得燃料组件的阻力特性曲线。在堆本体CFD分析模型强迫流动工况计算结果的基础上,开展了标准燃料组件自然循环数值模拟分析。计算结果表明,在设定工况下,不仅释热能安全载出,而且可保证热组件任何位置均不会发生冷却剂泡核沸腾和流动不稳定性。计算得到了自然循环建立过程组件内冷却剂温度、燃料包壳和芯体的温度分布、热点位置以及循环流量的变化规律,为研究热组件的瞬态热工水力特性提供了理论方法和参考数据。 相似文献
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针对一种新型的超临界水堆设计方案——混合能谱超临界水堆(SCWR-M)进行分析。混合能谱超临界水堆包括热谱区和快谱区两部分,分别布置在堆芯的外部与内部。它在继承了热谱与快谱超临界堆芯设计优点的同时,有效地克服了两者的不足。对于热谱区,冷却剂与慢化剂同向流动,大幅降低了燃料包壳的表面温度和组件的机械加工难度;对于快谱区,采用多层燃料组件和较大的栅距棒径比p/d,可得到较高的燃料转换比和较小的冷却剂负反应性系数。本工作采用自主开发的基于子通道分析和三维物理计算的耦合程序,对混合能谱超临界水堆的热工性能和中子物理性能(包括燃耗性能)进行研究。初步的耦合分析结果表明了混合能谱超临界水堆设计方案的可行性。 相似文献
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《Journal of Nuclear Science and Technology》2013,50(6):929-935
A supercritical water-cooled reactor (SCWR) was proposed as a kind of generation IV reactor in order to improve the efficiency of nuclear reactors. Although investigations on the thermal-hydraulic behavior in SCWR have attracted much attention, there is still a lack of CFD study on the heat transfer of supercritical water in fuel channels. In order to understand the thermal-hydraulic behavior of supercritical fluids in nuclear reactors, the local fluid flow and heat transfer of supercritical water in a 37-element fuel bundle has been studied numerically in this work. Results show that secondary flow appears and the cladding surface temperature (CST) is very nonuniform in the fuel bundle. The maximum cladding surface temperature (MaxCST), which is an important design parameter for SCWR, can be predicted and analyzed using the CFD method. Due to a very large circumferential temperature gradient in cladding surfaces of the fuel bundle, the precise cladding temperature distributions using the CFD method is highly recommended. 相似文献
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《Journal of Nuclear Science and Technology》2013,50(6):537-548
The feasibility of the sliding pressure startup of a high-temperature supercritical-pressure light water reactor (super LWR, SCLWR-H) is assessed from both thermal and stability considerations. In the sliding pressure startup, nuclear heating starts at subcritical pressure and the reactor is pressurized to supercritical pressure at a low power and high enough flow rate. The reactor power and flow rate are then raised gradually to the rated normal values at constant supercritical operating pressure. During startup, the maximum cladding surface temperature must not exceed 620°C. For two-phase flow at subcritical pressures, the homogeneous equilibrium model is used. The thermal-hydraulic and coupled neutronic thermal-hydraulic stabilities during pressurization and power-raising are investigated by a frequency-domain linear analysis for both supercritical-pressure and subcritical-pressure operating conditions. The same stability criteria as those of BWRs are used. From the analysis results, a sliding pressure startup procedure is proposed for super LWR. The thermal criteria are satisfied by keeping the core power between the maximum allowable limit and minimum limit required for turbine startup and operation. The thermal-hydraulic stability and coupled neutronic thermal-hydraulic stability can be maintained by applying an orifice pressure drop coefficient at the inlet of fuel assembly and by controlling the power and flow rate during startup. 相似文献
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Lanfranco Monti Jörg Starflinger Thomas Schulenberg 《Nuclear Engineering and Design》2011,241(5):1579-1591
The High Performance Light Water Reactor (HPLWR) is a thermal spectrum nuclear reactor cooled and moderated with light water operated at supercritical pressure. It is an innovative reactor concept, which requires developing and applying advanced analysis tools as described in the paper.The relevant water density reduction associated with the heat-up, together with the multi-pass core design, results in a pronounced coupling between neutronic and thermal-hydraulic analyses, which takes into account the strong natural influence of the in-core distribution of power generation and water properties. The neutron flux gradients within the multi-pass core, together with the pronounced dependence of water properties on the temperature, require to consider a fine spatial resolution in which the individual fuel pins are resolved to provide precise evaluation of the clad temperature, currently considered as one of the crucial design criteria. These goals have been achieved considering an advanced analysis method based on the usage of existing codes which have been coupled with developed interfaces.Initially neutronic and thermal-hydraulic full core calculations have been iterated until a consistent solution is found to determine the steady state full power condition of the HPLWR core. Results of few group neutronic analyses might be less reliable in case of HPLWR 3-pass core than for conventional LWRs because of considerable changes of the neutron spectrum within the core, hence 40 groups transport theory has been preferred to the usual 2 groups diffusion theory. Successively, with the usage of a developed pin-power reconstruction technique capable to account for the innovative fuel assembly design, sub-channel investigations of the individual fuel assemblies have been performed evaluating pin-wise clad temperatures. Obtained results will be discussed giving a detailed insight of the revolutionary HPLWR 3 pass core concept and understanding the physical reasons, which influence the local clad temperatures.The obtained results represent a new quality in core analyses, which takes into full consideration the coupling between neutronics and thermal-hydraulics as well as the spatial effects of the fuel assembly heterogeneity in determining the local pin-power and the associated maximum clad temperature. 相似文献
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Lead–alloy cooled fast reactor is one of the six Gen-IV reactors. It has many attractive features such as excellent natural circulation performance, better shielding against gamma rays or energetic neutrons and potentially reduced capital costs. A natural circulation lead–alloy cooled fast reactor with 10 MWth is under design in China (hereafter called LFR-10MW). Fuel assemblies thermal hydraulic analysis is of vital importance for a successful design. A subchannel analysis code with flow distribution model was used to carry out the thermal hydraulic analysis. This work briefly gave the thermal-hydraulic design for the LFR-10MW and analyzed the thermal-hydraulic characteristics under steady-state condition using the subchannel analysis code. Whole core analysis was performed to locate the hottest fuel assembly using the code. The hottest fuel assembly was analyzed to obtain the cladding temperature, fuel temperature and coolant velocity. The maximum cladding temperature, the maximum fuel center temperature and the maximum coolant velocity are all below the design constraints. These results imply that the thermal-hydraulic design of LFR-10MW is feasible. 相似文献
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超临界水冷堆燃料验证实验(SCWR-FQT)将对1个小型燃料组件在超临界水环境下进行堆内性能测试。为了对该实验回路进行系统设计和安全分析,应用修改过的ATHLET程序建立实验回路计算模型,对两种造成燃料组件实验段冷却剂流量部分或全部丧失的设计基准事故进行模拟分析,即由于装载实验段的压力管内部的导向管破裂导致流经实验段的冷却剂旁通和主冷却剂泵卡轴事故。计算结果显示:实验段冷却剂旁通事故中,燃料包壳温度在事故初期出现约920 ℃的峰值;而主泵卡轴事故中,燃料包壳温度未明显升高。计算结果表明,现有的安全系统设计能保证在事故情况下维持燃料组件实验段的有效冷却。 相似文献
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This paper describes study on the procedure of raising the reactor thermal power and the reactor coolant flow rate during the power-raising phase of plant startup for the supercritical water-cooled fast reactor (SWFR), which is selected as one of the Generation IV reactor concepts. Since part of the seed fuel assemblies and all the blanket fuel assemblies of the SWFR are cooled by downward flow, the feedwater from the reactor vessel inlet nozzle to the mixing plenum located below the core is distributed among these fuel assemblies and the downcomer. The flow rate distribution as the function of both the reactor thermal power and the feedwater flow rate, which are the design parameters for the power-raising phase, is obtained by the thermal hydraulic calculations. Based on the flow rate distribution, thermal analyses and thermal-hydraulic stability analyses are carried out in order to obtain the available region of the reactor thermal power and the feedwater flow rate for the power-raising phase. The criteria for the “available” region are the maximum cladding surface temperature (MCST) and the decay ratio of thermal-hydraulic stability in three “hot” channels; two seed assemblies with upward/downward flow and a blanket assembly. The effects of various heat transfer correlations and axial power distributions are also studied. 相似文献