超临界水冷堆CSR1000燃料组件子通道分析

利用实验数据和计算流体力学(CFD)商用软件CFX对现有子通道分析模型进行研究,分析其在超临界水冷堆(SCWR)分析中的适用性,并根据分析结果对ATHAS程序进行改进。采用改进的ATHAS程序对超临界水冷堆CSR1000燃料组件进行稳态子通道分析,获得燃料组件冷却剂和包壳温度分布、流动压降等参数。结果表明:减小螺旋肋螺距(Hw)可展平燃料组件冷却剂出口温度分布、降低包壳表面最高温度(MCST),但同时燃料组件流动阻力将增大。     

超临界水冷混合堆快谱区组件物理——热工分析

提出了超临界水冷混合堆快谱区多层燃料组件设计方案.应用MCNP程序为该组件建立计算模型,并进行了相应的物理计算;同时运用子通道分析程序STAFAS对多层燃料组件子通道进行了初步的稳态热工分析.计算结果表明:超临界水冷混合堆快谱区多层燃料组件燃料转换比超过1.0,并且获得负的冷却剂空泡反应性系数;燃料包壳表面最高温度约为595℃,低于设计准则规定的上限值,同时组件各子通道出口冷却剂温度均匀性较好.通过对燃料棒径敏感性分析可知,较大棒径组件燃料转换比较大,但也会导致热通道包壳表面温度峰值升高.     

超临界水冷堆双排棒组件子通道分析

首先利用先进子通道分析程序(ATHAS)对超临界水冷堆(CGN-SCWR)的双排棒组件进行子通道分析,以考察燃料棒包壳温度等热工参数是否达到安全要求。根据分析结果结合子通道水力直径和冷却剂出口温度,选取一些典型子通道的热工参数结果做详细比对,了解组件中不同类型子通道内的热工参数变化对组件性能的影响。另外,对子通道计算采用的湍流交混系数、轴向摩擦系数和传热关系式进行敏感性分析,以了解经验关系式对计算结果的影响。结果显示:所有热工参数结果均达到设计要求,包壳最高温度为685.3℃,且不同传热关系式的选择对包壳温度的影响明显,最大温差达到了41.3℃。     

行波堆TP-1堆芯热工水力单通道与子通道分析方法研究

以泰拉能源公司提出的钠冷行波堆TP-1为研究对象,通过钠冷行波堆瞬态安全分析程序TAST得到堆芯各组件内冷却剂、包壳和燃料棒的平均温度分布。用子通道分析程序SACOS-Na对TAST计算得到的最热组件进行详细分析计算,得到该组件内冷却剂的温度、压力和流速分布,并得到燃料棒和包壳的温度场。结果表明：单通道与子通道的结合使用能有效提高计算效率,提高反应堆设计的安全性。     

CSR1000堆芯流量分配的研究

以中国百万千瓦级超临界水冷堆(CSR1000)堆芯为研究对象,建立热工水力计算模型,计算出冷却剂和慢化剂温度分布、堆芯功率分布、燃料组件出口压力及流量分配等参数。计算结果表明,适当增加堆芯内部燃料组件流量比例,可以有利于径向功率展平,内外燃料组件通道出口压降,呈现"N"型变化,增大内部燃料组件的堆芯入口功率,内部组件内的流量分配也将减少,而外部燃料组件通道中的流量将增加,适当调整堆芯入口流量初始分配比例,可以使各通道功率分布展平。     

中国先进研究堆标准燃料组件三维流场分析

采用计算流体力学软件CFX4.4和CFX5.5对中国先进研究堆标准燃料组件流场进行了数值模拟。计算得到了额定工况下标准燃料组件内各个冷却剂通道的流量分布和不等间隙通道燃料板两侧压差。根据不同流量下的压降计算结果,给出了标准燃料组件的阻力特性曲线,并与试验结果进行了比较,符合较好。     

VVER反应堆燃料组件流动传热特性CFD分析

采用计算流体力学（CFD）方法对俄罗斯水-水高能反应堆（VVER）先进燃料组件（AFA）的流动传热特性进行模拟,获得了额定工况下燃料组件冷却剂流场、流动压降和温度分布等。结果表明:与内部含交混翼的格架相比,AFA燃料组件定位格架的压力损失较小;定位格架围板导向翼附近存在滞流现象,导致燃料组件外围区域冷却剂温度偏高;不同的测量管周向棒功率比K_c对燃料组件出口冷却剂温度的测量值有较大影响。该分析结果可为核电厂堆芯温升预警值ΔT_t的设定提供参考。      

海洋条件下反应堆热工水力参数的子通道计算

针对海洋条件下反应堆的子通道热工水力分析,建立了海洋运动附加力模型和瞬态入口边界,将起伏、摇摆及复合运动的附加力关系式用于子通道模型的轴向和横向动量方程,并应用到COBRAⅢC程序将其改造为适应海洋条件的反应堆子通道分析程序。作为验证,计算了加热实验通道和"奥陆"堆在起伏运动情况下热通道的临界热流密度比(CHFR)、出口空泡份额和冷却剂流量,并与文献结果对比。还详细计算了"奥陆"堆在起伏、不同摇摆中心和复合运动情况下,热通道的CHFR和不同位置子通道出口的热工水力参数。研究表明:海洋条件下反应堆的子通道热工水力参数随运动呈周期性变化;起伏运动对子通道的压降影响较大,摇摆运动对子通道冷却剂的流量和温度影响较大。     

超临界水冷动力堆燃料组件的子通道研究分析

与目前的轻水堆相比较，由于超临界水冷动力反应堆(SCPR)的热效率高、反应堆系统简单，预计将降低发电成本高热效率通过超临界压力水冷却来获得、如果冷却剂流体在燃料组件中的分布是非均匀的．由于冷却剂温度提高、冷却剂密度的变化而出现大的流量偏移和传热系数降低的复合效应，燃料包壳的表面温度会局部升高：因此，SCPR燃料组件设计采用基于沸水堆的SILFEED的子通道分析程序SCPR燃料组件具有许多正方形水棒、燃料棒被布置在这些水捧周围。燃料棒的间距和直径分别为11．2nun和10．2mm。由于冷却剂流体在燃料组件内的分布主要取决于燃料棒和水棒之间的间隙宽度、对适当的间隙宽度进行了研究。子通道分析表明，在间隙宽度为1．0mm时，冷却剂流量分布是均匀的，最高的燃料包壳表面温度低于600℃、在设计中提高了燃料包壳的温度裕度。     

铅铋冷却快堆含绕丝燃料组件子通道程序开发与验证

由于铅铋冷却剂流动传热现象的复杂性,准确计算铅铋冷却含绕丝燃料组件的冷却剂和包壳温度是液态金属冷却快堆燃料组件热工分析的重点。本文基于集总参数法对守恒方程进行求解,开发了适用于铅铋冷却快堆的子通道分析程序,对液态铅铋在棒束燃料组件中的摩擦阻力模型、湍流交混模型和对流换热模型进行了适用性分析,并对7棒束大涡模拟和19棒束含绕丝传热实验进行了对比验证。结果表明：包壳和冷却剂温度的最大相对误差低于5%。程序能较好完成铅铋冷却含绕丝燃料组件的热工水力计算,可为铅铋冷却快堆设计提供支持。     

International Solvent Extraction Conference ISEC ′90

The supercritical-water-cooled power reactor (SCPR) is expected to reduce power costs compared with those of current LWRs because of its high thermal efficiency and simple reactor system. The high thermal efficiency is obtained by supercritical pressure water cooling. The fuel cladding surface temperature increases locally due to a synergistic effect from the increased coolant temperature, the expanded flow deflection due to coolant density change and the decreased heat transfer coefficient, if the coolant flow distribution is non-uniform in the fuel assembly. Therefore, the SCPR fuel assembly is designed using a subchannel analysis code based on the SILFEED code for BWRs.

The SCPR fuel assembly has many square-shaped water rods. The fuel rods are arranged around these water rods. The fuel rod pitch and diameter are 11.2 mm and 10.2 mm, respectively. Since coolant flow distribution in the fuel assembly strongly depends on the gap width between the fuel rod and the water rod, the proper gap width is examined. Subchannel analysis shows that the coolant flow distribution becomes uniform when the gap width is 1.0 mm. The maximum fuel cladding surface temperature is lower than 600°C and the temperature margin of the fuel cladding is increased in the design.     

A comparative study of the MATRA-LMR-FB calculation with the SABRE result for the flow blockage accident in the sodium cooled fast reactor

A sub-channel flow blockage may be initiated by an ingression of damaged fuel debris or foreign obstacles into a core subassembly for the sodium cooled fast reactor (SFR) due to the compact design of the fuel arrangement. Since local coolant temperature could go up high enough to reach a safety limit by the blockage disturbance in the subassembly, the MATRA-LMR-FB code was developed to analyze such blockage effect. An effort has been undergoing to enhance its reliability.In this study, a code-to-code comparison analysis with another code, SABRE4, was performed to supplement a qualification of the MATRA-LMR-FB. The two codes were applied to the analysis of partial sub-channel blockage accidents in a subassembly of the KALIMER-150, which is a conceptual design of a sodium-cooled fast reactor with an electric output of 150 MW. The analyses were carried out not only for radially different blockage positions but also for different blockage sizes in the subassembly.In result, the two code results were generally agreed both in magnitude and trend within a range. Therefore, it was concluded that the comparison results could support complementarily the applicability of the MATRA-LMR-FB to the partial flow blockage accident in the subassembly of the SFR.     

Subassembly experiments and a computer code to analyze the dynamic core deformation during local failure propagation

Safety investigations for LMFBRs have to consider local failure situations in one fuel element which may escalate to a hypothetical CDA. Such initiating events could produce high pressure pulses in a single subassembly which may expand and rupture the wrapper as well as load adjacent elements impulsively. The associated nonlinear dynamic core deformation problem is treated in this paper. In particular the multirow structural dynamics code CØRE-1 and underlying mechanical models are described. Each subassembly is simulated by an equivalent system of point masses and nonlinear coupling springs. The motion of the coolant layer between the elements is treated by an incompressible, non-stationary frictional flow model. In order to obtain realistic code input four types of static single subassembly deformation experiments are described which provided strongly nonlinear load deformation characteristics. Furthermore the transient pressure distribution within the core is obtained from a full scale explosion test. Finally code application is demonstrated and results are given of a transient analysis of the SNR 300 core.     

高温气冷堆热工分析模型改进与堆芯温度场分析

更准确地模拟球床式高温气冷堆堆芯温度分布,是反应堆安全分析尤其是超高温运行研究中的关键问题之一。由于堆芯球流运动具有不确定性,石墨块和碳砖等结构材料采用散体布置,堆内冷却剂流道复杂,对热工水力准确模拟造成困难,可进一步优化。本文结合HTR 10的结构特点和流道特征,简要分析了堆芯传热过程,说明了在热工模拟中准确划分结构和流道对获取更精确的堆芯温度分布的重要意义。详细梳理了冷却剂流动路径,改进了在THERMIX程序下建立的HTR 10原有热工分析模型,更合理地模拟了堆芯冷却剂漏流行为,使得模型对堆芯冷却剂流动和传热过程的描述更准确。与试验数据对比,改进后的模型对堆芯外围系统的温度分布模拟准确性显著提升。计算结果表明,反应堆在额定设计工况下满功率稳态运行时,燃料和反射层最高温度均未超过材料的耐热限值。     

快堆组件数值模拟研究

为详细研究快堆组件棒束中的流动换热特性,本工作采用Fluent程序对169棒束快堆燃料组件进行三维数值模拟。结果表明,在流量为10.92~18.67 kg/s时,计算得到的压降与已公开发表文献结果的相对偏差小于3.41%。内子通道的相对温度升高,呈现出周期为1/3螺距的波动,内子通道的局部温度比子通道程序SUPERENERGY计算的结果更高。根据模拟计算结果可更为准确地预测棒束通道内的流动换热情况,为今后组件棒束热工水力学设计提供参考。     

数值反应堆堆芯通道级三维热工水力程序CorTAF开发及初步验证

堆芯是核动力系统的核心部件,其完整性是反应堆安全运行的重要前提。传统核反应堆堆芯热工水力分析方法无法满足未来先进核动力系统的高精度模拟需求。本文依托开源CFD平台OpenFOAM,针对压水堆堆芯棒束结构特点建立了冷却剂流动换热模型、燃料棒导热模型和耦合换热模型,开发了一套基于有限体积法的压水堆全堆芯通道级热工水力特性分析程序CorTAF。选取GE3×3、Weiss和PNL2×6燃料组件流动换热实验开展模型验证,计算结果与实验数据基本符合,表明该程序适用于棒束燃料组件内冷却剂流动换热特性预测。本工作对压水堆堆芯安全分析工具开发具有参考和借鉴意义。     

Analysis of three different types of blockage in a sodium flow path with the MATRA-LMR-FB code

A local blockage in a subassembly of an LMR is of particular importance because the local temperature of the coolant increases at the downstream of a blockage and the integrity of the fuel clad can be threatened when an obstacle or a blockage is formed in a flow path. To analyze a flow blockage in a nuclear reactor core, Korea Atomic Energy Research Institute developed a subchannel analysis computer code MATRA-LMR-FB. This code adopts several enhanced modeling features such as a distributed resistance model, state-of-the-art turbulent mixing models, a hybrid difference scheme, and a porous body pressure drop model, therefore, it is applicable to a flow path with a plate-type or a porous-type blockage. The effect of each model has been evaluated through an analysis for the THORS experiment, in which a plate-type blockage was located in a flow channel with wire-wrapped fuel rods. The overall capability of the code has also been evaluated for the KNS experiment with a plate-type blockage in a grid-spaced flow channel, and for the SCARLET-II experiments with a porous blockage in channels formed by wire-wrapped fuel rods. The code shows good predictions for the experiments with a wire-wrapped flow path with a plate-type or a porous type blockage. The analyses for the KNS experiments reveal that the code requires a precise blockage model related to a grid spacer model.     

某游泳池式反应堆全厂断电事故分析

采用RELAP5/MOD3.3程序对某游泳池式反应堆的全厂断电事故工况进行计算,对堆内冷却剂流动逆转过程进行了模拟计算,并对全厂断电事故下堆芯漏流和组件间流动等相关参数对流动逆转的影响进行了深入分析。结果表明,该反应堆在失去全部强迫流动的情况下,能形成足够的自然循环流量,以导出堆芯余热,燃料组件不会发生破损。