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。     

PARCS稳态两相热工水力程序PATHS的开发与耦合

本工作开发了PARCS的先进热工水力求解器PATHS,可对沸水堆进行热工水力稳态模拟。与RELAP5的计算结果进行验证,结果表明,PATHS的计算结果与RELAP5的基本一致。将PATHS与PARCS进行耦合,对SMART反应堆及Peach Bottom 2 OECD Turbine Trip基准题进行计算,结果表明,PARCS/PATHS耦合程序计算结果准确有效,能用于沸水堆的稳态物理热工耦合计算。     

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。     

PT-SCWR燃料组件的物理热工耦合分析

利用蒙特卡罗程序(MCNP)和子通道程序ATHAS对压力管式超临界水堆(PT-SCWR)燃料组件进行物理热工耦合分析;这种耦合方式是合理有效的。分析结果表明:PT-SCWR组件中燃料富集度的分布对燃料组件的径向功率分布有很大影响,通过调节各圈棒束的燃料富集度,可以有效地改善径向功率分布;慢化剂厚度对棒束轴向功率分布有明显影响,当慢化剂厚度为25 cm时,轴向功率分布最接近余弦形状。     

超临界水堆铀钍混合燃料组件中子学特性分析

针对超临界水堆的能谱特性及钍燃料的中子特性,提出了一种应用于超临界水堆的新型铀钍混合燃料组件设计方案,并利用组件计算程序Dragon"对该设计在不同工况下的中子学特性进行了分析,包括:无限增殖因数、反应性温度系数、易裂变材料存量比(FIR)等,以及它们随燃耗变化的规律。另外,通过改变混合燃料组件中燃料棒的慢化剂-燃料比,探究了其对燃料组件中子学特性的影响。结果表明:超临界水堆较硬的中子能谱有利于产生易裂变核素,同时该新型燃料组件在提高燃料利用率和减少次锕系元素存量方面具有一定的优势。     

超临界水冷堆系统分析程序开发

详细介绍了自主开发的超临界水堆(SCWR)安全分析程序SCTRAN的数学模型、辅助方程及计算流程。运用圆管内超临界水的喷放实验数据和西屋公司SCWR大破口失水事故(LOCA)数据对SCTRAN程序的有效性进行验证。验证结果表明,SCTRAN计算结果与程序APROS基本一致,对西屋公司SCWR非能动冷却剂系统的事故分析结果同RELAP5-3D程序的结果基本一致,计算结果可靠性较高,具备对SCWR进行事故分析的能力。     

RELAP5堆芯功率计算模型的扩展

为了更好地将反应堆热工水力最佳估算程序RELAP5应用于分析控制棒控制的反应堆堆芯的功率瞬变过程，堆芯功率计算模块除保留原程序中使用的点堆中子动力学模型外，还必须向轴向一维中子动力学模型进行扩展。本文通过在现有轴向一维物理程序基础上进行改造和开发，实现了RELAP5程序与一维物理程序的耦合，并且通过例题验证了耦合的正确性。     

混合能谱超临界水堆失流事故缓解措施研究

使用改进的系统程序RELAP5建立了一个混合能谱超临界水堆(SCWR-M)模型。为研究混合能谱超临界水堆失流事故特性,以获取缓解混合能谱超临界水堆失流事故的措施,选取反应堆冷却剂泵惰转时间、压力容器上部储水空间容积和安注流量作为主要参数进行分析。研究表明,混合能谱超临界水堆系统的设计是可行的。反应堆冷却剂泵惰转15 s,压力容器上部水空间容积大于27 m³,以及安注流量高于系统满功率稳态流量的5%是缓解混合能谱超临界水堆失流事故的主要措施。     

超临界水冷堆堆芯子通道稳态热工分析

超临界水冷堆(SCWR)作为6种第四代未来堆型中唯一的水冷堆,冷却剂出口温度可达500℃,具有良好的经济性.本文采用改进的COBRA-IV程序对超临界水冷堆方形组件子通道进行稳态热工分析.对计算结果进行分析可知:减小慢化剂通道中给水质量流量份额和加大慢化剂通道与相邻子通道之间的热阻,可以降低热管焓升,后者还可以得到较好的慢化效果.通过热通道的传热恶化分析发现,超临界水冷堆的设计不能避免传热恶化,必须精确计算传热恶化条件下的包壳温度才能确定包壳能否保证其完整性.     

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。     

超临界水冷堆的安全分析

应用RELAP5-3D程序建立了超临界水冷堆(SCWR)的稳态模型,并在此基础上,分别对SCWR的两种瞬态和两种事故工况进行了分析。汽轮机旁路系统的存在可有效维持反应堆压力,保证反应堆安全。若SCWR失去给水,在辅助给水系统启动之前,向下流的水棒可通过热传导带走堆芯热量,并向燃料通道内提供冷却剂,缓解堆芯升温。因而,向下流的水棒体现了SCWR的安全性。主泵卡轴事故由于没有惰转,最热包壳温度值最大,因而主泵惰转可有效缓解包壳温度的升高。     

基于SCWR堆芯结构的子通道程序开发与应用

为能够对超临界水堆（SCWR）堆芯进行子通道分析,开发了新的子通道分析程序SABER。该程序在COBRA程序的基础上改进了网格结构和热传导模型,加入了新的边界条件和水物性模块,以适用于SCWR慢谱燃料组件的子通道分析。为评估程序的适用性,采用该程序对SCWR堆芯概念设计中的慢谱燃料组件进行子通道建模,并进行稳态计算。结果表明,该程序能够用于SCWR堆芯的子通道计算分析,并较好地解决了慢谱组件计算中慢化通道和冷却通道间的热耦合及逆向流动的模拟问题。     

Studies on supercritical water reactor fuel assemblies using the sub-channel code COBRA-EN

In the Generation IV International Forum (GIF) program, the supercritical water reactor (SCWR) concept is among the six innovative reactor types selected for development in the near future. In principle the higher efficiency and better economics make the SCWR concept competitive with the current reactor design. Due to different technical challenges that, however exist, fuel assembly design represents a crucial aspect for the success of this concept. In particular large density variations, low moderation, heat transfer enhancement and deterioration have a strong effect on the core design parameters. Only a few computational tools are currently able to perform sub-channel thermal-hydraulic analysis under supercritical water conditions. At JRC-IE the existing sub-channel code COBRA-EN has been improved to work above the critical pressure of water. The water properties package of the IAPWS Industrial Formulation 1997 was integrated in COBRA-EN to compute the Thermodynamic Properties of Water and Steam. New heat transfer and pressure drop correlations more indicated for the supercritical region of water have also been incorporated in the code. As part of the efforts to appraise the new code capabilities, a code assessment was carried out on the hexagonal fuel assembly of a fast supercritical water reactor. COBRA-EN was also applied in combination with the neutronic code MCNP to investigate on the use of hydride fuel in the HPLWR supercritical water fuel assembly. The results showed that COBRA-EN was able to reproduce the results of similar studies with acceptable accuracy. Future activities will focus on the validation of the code against experimental data and the implementation of new features (counter-current moderator channel, wall, and wire-wrap models).     

Thermal-hydraulic and neutron-physical characteristics of a new SCWR fuel assembly

A new fuel assembly design for a thermal supercritical water cooled reactor (SCWR) core is proposed. Compared to the existing fuel assemblies, the present fuel assembly has two-rows of fuel rods between the moderator channels, to achieve a more uniform moderation for all fuel rod cells, and subsequently, a more uniform radial power distribution. In addition, a neutron-kinetics/thermal-hydraulics coupling method is developed, to analyze the neutron-physical and thermal-hydraulic behavior of the fuel assembly designs. This coupling method is based on the sub-channel analysis code COBRA-IV for thermal-hydraulics and the neutron-kinetics code SKETCH-N for neutron-physics. Both the COBRA-IV code and the SKETCH-N code are accordingly modified. An interface is established for the data transfer between these two codes. This coupling method is applied to both the one-row fuel assemblies (previous design) and the two-row fuel assemblies (present design). The performance of the two types of fuel assemblies is compared. The results show clearly that the two-row fuel assembly has more favorable neutron-physical and thermal-hydraulic characteristics than the one-row fuel assembly. The effect of various parameters on the fuel assembly performance is discussed. The coupling method is proven to be well suitable for further applications to SCWR fuel assembly design analysis.     

超临界水冷堆双排燃料组件子通道分析

研究基于Cobra-IV程序,开发了适用于超临界水冷堆燃料组件分析的子通道程序.针对超临界水冷堆慢谱双排组件,进行了稳态计算,获取了相关组件热工水力参数.在此基础上,针对单一通道进行了瞬态计算,分析了燃料棒线功率变化和冷却剂流量变化条件下,超临界水冷堆燃料组件的流动和传热的动态响应,为超临界水冷堆组件的优化设计提供了参考.     

ATHLET-SC程序的开发及适用性分析

由于超临界水堆（SCWR）在系统简化、降低成本和提高热效率上的优势，SCWR的研究在全球范围内得到广泛关注。在众多有关超临界水堆的研发工作中，开发适用于SCWR的系统分析程序是进行SCWR系统设计和安全评估的关键技术难题之一。本工作基于最佳估算系统分析程序ATHLET2.1A，增加了超临界热物性参数，开发出适用于SCWR的系统分析程序ATHLET-SC，将现有的ATHLET程序扩展到超临界压力状态。为评估修改后的程序的适用性，建立了混合能谱超临界水堆堆芯模型，并对该模型进行了功率瞬态计算。此外，对1个简化的超临界水冷却回路进行了稳定性分析。计算结果表明：修改过的ATHLET程序（ATHLET-SC）对SCWR系统的模拟具有良好的适用性。     

Safety analysis of reactivity insertion accidents in a heavy water nuclear research reactor core using coupled 3D neutron kinetics thermal-hydraulic system code technique

Nuclear power plant Safety analysis using coupled 3D neutron kinetics/thermal-hydraulic codes technique is increasingly used nowadays. Actually, the use of this technique allows getting less conservatism and more realistic simulations of the physical phenomena. The challenge today is oriented toward the application of this technique to the operating conditions of nuclear research reactors. In the current study, a three-Dimensional Neutron Kinetics and best estimate Thermal-Hydraulic model based upon the coupled PARCS/RELAP5 codes has been developed and applied for a heavy water research reactor. The objective is to perform safety analysis related to design accidents of this reactor types. In the current study two positive reactivity insertion transients are considered, SCRAM protected and self-limiting power excursion cases. The results of the steady state calculations were compared with results obtained from conventional diffusion codes, while transient calculations were assessed using the point kinetic model of the RELAP5 code. Through this study, the applicability and the suitability of using the coupled code technique with respect to the classical models are emphasized and discussed.