混合能谱超临界水堆堆芯热工-物理性能分析

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。     

混合能谱超临界水堆堆芯设计分析

提出了一种新型的超临界水堆概念设计:混合能谱超临界水堆,它包括慢谱区和快谱区两部分.其慢谱区燃料组件采用双排燃料组件,快谱区采用简单的正方形栅元燃料组件.慢谱区与快谱区的燃料组件都采用同向流动方式来简化堆芯设计.慢谱区的冷却剂出口温度远低于整个堆芯的出口温度,这大大降低了慢谱区包壳的温度峰值.此外,由于快谱区冷却剂密度很小,流速很高,故可采用较大的栅元结构,这有效地降低了包壳周向局部传热不均匀性.所以混合堆在充分继承慢谱、快谱堆芯优点的基础上,弥补两者的不足.     

超临界水混合堆快谱区组件物理-热工性能分析

提出超临界水混合堆快谱区多层燃料组件设计方案。用MCNP与STAFAS程序对多层燃料组件进行初步的中子物理与热工水力性能分析，同时对组件结构参数（栅距与棒径比P/D）进行敏感性研究。结果表明：快谱多层燃料组件设计不仅能够实现核燃料的增殖，且可获得较大的负冷却剂温度反应性系数与燃料温度反应性系数;减小P/D均可提高燃料的转换比，但较小P/D会导致核热点因子增大。适当调整组件裂变区燃料富集度可有效改善组件裂变区轴向功率不均匀性，降低核热点因子。     

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。     

燃料棒周向导热对超临界水堆堆芯子通道分析的影响

本文在子通道程序的燃料棒模型中引入三维导热方程,使该模型能用来模拟燃料棒的周向导热情况。采用改造后的子通道程序对混合谱超临界水堆设计中的两种燃料组件结构进行计算分析,研究燃料棒周向导热对超临界水堆燃料组件子通道分析的影响。结果表明：热谱组件的子通道计算中,燃料棒周向导热的影响不能忽略;快谱组件的子通道计算中,燃料棒周向导热的影响基本可忽略。     

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。     

混合能谱超临界水堆快谱组件优化设计

本工作从热工水力和中子物理两方面对混合能谱超临界水堆混合谱堆芯的快谱区多层组件进行优化设计。对于轴向以再生区和裂变区交替布置的快谱组件,分别改变其轴向布置方式、燃料芯块直径、栅径比及外围燃料棒距组件盒最小距离,并分析它们对组件热工和物理性能的影响,从而得到较优的参数范围,尽可能提高混合谱超临界水堆的固有安全性和经济性。     

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。     

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。     

超临界水堆子通道分析

超临界水堆作为6种第4代未来堆型中唯一的水冷堆，具有一些独特的特点，受到了广泛重视。本工作以上海核工程研究设计院的常规压水堆子通道程序为基础，开发编制了适用于超临界水堆的子通道程序，并对典型带有慢化剂水棒的超临界水堆燃料组件进行了模拟计算，得到了堆芯子通道内的温度、燃料棒包壳温度、表面传热系数等参数的分布规律。此外，研究了不同超临界流体换热关系式对计算结果的影响，结果显示，各传热关系式的计算结果存在一定差异。     

Core and sub-channel analysis of SCWR with mixed spectrum core

The SCWR core concept SCWR-M is proposed based on a mixed spectrum and consists of a thermal zone and a fast zone. This core design combines the merits of both thermal and fast SCWR cores, and minimizes their shortcomings. In the thermal zone co-current flow mode is applied with an exit temperature slightly over the pseudo-critical point. The downward flow in the thermal fuel assembly will provide an effective cooling of the fuel rods. In the forthcoming fast zone, a sufficiently large negative coolant void reactivity coefficient and high conversion ratio can be achieved by the axial multi-layer arrangement of fuel rods. Due to the high coolant inlet temperature over the pseudo-critical point, the heat transfer deterioration phenomenon will be eliminated in this fast spectrum zone. And the low water density in the fast zone enables a hard neutron spectrum, also with a wide lattice structure, which minimizes the effect of non-uniformity of the circumferential heat transfer and reduces the cladding peak temperature.     

Coupled thermal-hydraulics and neutron-physics analysis of SCWR with mixed spectrum core

The new SCWR conceptual design (SCWR-M) is proposed on the basis of a mixed spectrum core consisting of a thermal spectrum zone and a fast spectrum zone. This new core design is considered to be the hybrid of the existing thermal SCWR and fast SCWR cores. It combines the merits of both thermal and fast SCWR cores, at the same time minimizes their shortcomings. For the thermal zone, the difficulties in the mechanical design and the maximum cladding temperature can be reduced as far as possible by the co-current flow mode; and for the fast zone, a sufficiently large negative coolant void reactivity coefficient and breeding ratio can be achieved by the multi-layer arrangement of fuel rods.The performance, including the burn-up behavior, of the proposed core is investigated with 3-D coupled neutron-physical and thermal-hydraulic calculations. During the coupling procedure, the thermal-hydraulic behavior is analyzed using a sub-channel analysis code and the neutron-physical performance is computed with a 3-D diffusion code. The results obtained so far have shown that the mixed spectrum SCWR concept (SCWR-M) is feasible and promising.     

Subchannel analysis of fuel assemblies of a lead–alloy cooled fast reactor

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.     

SCWR-FQT回路的冷却剂流量丧失事故研究

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。     

Accuracy of nuclear design of fast and thermal neutron coupled core by SRAC

Supercritical-pressure light water cooled fast reactor adopts the blanket fuel assemblies with depleted uranium fuel and zirconium hydride layer in the core for negative coolant void reactivity. Thermal neutrons are generated in the core of fast reactor. It is called “fast and thermal neutron coupled core”. The purpose of the present study is to examine the accuracy of assembly and core calculations including preparation of the macroscopic cross sections with the SRAC code system for “fast and thermal neutron coupled core” in comparison with the Monte Carlo codes, MVP and MVP-BURN. Accuracy of the neutron multiplication factor and coolant void reactivity calculation has been evaluated in four types of cores of different fractions of the blanket fuel assembly with zirconium hydride rods. The conventional analysis is based on the macroscopic cross sections which are prepared with infinite lattice. The conventional SRAC calculation underestimates the neuron multiplication factor for all types of cores. Other findings are that the conventional SRAC calculation overestimates coolant void reactivity for the cores without zirconium hydride rods, and underestimates coolant void reactivity for the core of all blanket fuel assemblies having zirconium hydride rods. To overcome these problems, it has been proposed that the macroscopic cross sections of seed fuel assembly are prepared with the model that a seed fuel assembly is surrounded by blanket fuel assemblies in order to take into account the effects of the surrounding fuel assemblies. Evaluations show that accuracy of the neutron multiplication factor by the SRAC calculation can be improved by the proposed method.     

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。