日本文殊快堆整体热腔室流场三维数值模拟

为研究日本文殊快堆一回路热腔室的热工水力特性,借鉴和消化国外快堆的设计经验,使用流体力学软件CFX对文殊快堆整体热腔室进行三维稳态数值模拟,得到其整体热腔室流场。文殊快堆全堆芯温度监测系统可为我国快堆小型化设计作技术准备。     

日本文殊原型快堆堆芯出口腔室热分层现象数值模拟

本文利用商业CFD程序STAR-CCM+,采用合理的网格生成技术及物理模型,对日本文殊原型快堆堆芯出口腔室建立近似1∶1的模型,模拟分析40%额定功率停堆过程中堆芯出口腔室的瞬态工况,获得腔室内较为完整的热分层进程。结果表明：停堆2 min后腔室内出现稳定热分层现象;10～21 min时热分层通过上升桶桶顶位置;10～140 min热分层处于上升筒顶端位置附近期间,腔室内流型不稳定;140 min后热分层完全处于上升桶顶,桶内流型稳定且接近于停堆前。模拟结果与实验数据对比表明,停堆初期4 min内两者符合较好,表明本文模拟方法适用于停堆工况堆芯出口腔室热分层进程模拟;之后模拟进程明显快于实验,分析其偏差主要来自模拟边界及结构与实际的差异。     

日本文殊堆启动对快堆研究至关重要

【本刊2003年11月综合报道】 2003年3月28日,日本核燃料循环开发机构(JNC)执行副总裁Yasuo Nakagami、法国原子能委员会(CEA)核能部主任Jacques Bouchard和美国阿贡国家实验室主任Hermann A.Grunder在美国发表了一份联合声明,强烈支持尽快恢复日本文殊原型快堆的运行。该声明认识到作为下一代核系统的快堆及其配套燃料循环的重要性,并强调文殊堆未来将扮演的重要角色。该堆的运行许可证于2003年1月底被日本最高法院吊销。 上述核工业界领导人强调,美国、日本和几个欧洲国家多年来通过建造实验型和原型快堆已获取了相当可观的专门知识…     

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。     

中国实验快堆事故停堆后余热排放过程的数值模拟

分析了中国实验快堆事故停堆后余热的排放过程。对热钠池中的流动与传热采用多孔介质模型的全三维数值模拟,对堆芯支路、事故热交换冷却回路和空冷塔冷却支路采用一维系统分析程序进行数值模拟。通过三维部分和一维部分相互耦合,模拟了余热排放的瞬态过程,得到了堆芯出口温度、燃料元件包壳的最高温度、余热热交换器的余热排放功率等许多重要参数随时间的变化曲线,对中国实验快堆的安全设计有重要的参考价值     

基于中国实验快堆的居里点非能动停堆系统的工作温度范围研究

非能动停堆装置可以大大提高钠冷快堆在无保护瞬态事故下的安全性,开展相关研究是十分必要的。采用居里点磁性合金的自动作停堆装置是目前国际上研究的主流装置之一。本文基于中国实验快堆(CEFR)的基本参数对采用居里点磁性合金的自动作停堆装置(Self-Actuated Shutdown System,简称SASS)的居里点温度预设值进行了研究。利用三维CFD程序采用大涡模拟的方法对安全棒附近的出口钠温进行计算分析,得到了温度振荡的幅度和频率,从而估算出居里点磁性材料正常工作的温度范围,确定了居里点温度预设值的下限。采用系统分析程序针对CEFR的无保护失流事故和无保护超功率事故进行分析,对居里点温度预设值的上限进行了评估,综合得出了居里点温度的预设值范围。本文通过以上工作,得出了一套居里点温度预设值的确定方法,对池式钠冷快堆的非能动停堆系统设计具有一定的指导意义。     

日本“文殊”快堆预运行试验开始

【《日本原子》1991年5月号第4页报道】日本动力堆核燃料开发事业团(PNC)耗资约600亿日元建造的文殊原型快堆已于1991年5月18日开始进行预运行试验。这座快堆充分利用了日本自己的技术和通过国际合作得到的资料。“文殊”是日本的第一座发电快堆,受到了日本国内外人士的关注。这次预运行试验将持续到8月份。     

快堆全堆芯热工流体子通道并行模拟技术研究

采用简化堆芯模型的传统子通道模拟计算结果难以精确反映堆芯的真实运行状况,利用高性能计算技术进行全堆芯精确到每个真实流道的子通道模拟计算成为研究热点。本文抽象描述了快堆堆芯的基础几何结构,在此基础上提出了一种全堆芯子通道建模方法和一种自适应的并行任务划分方法。设计了广度优先划分算法和层次划分算法,实现了全堆芯子通道任意个数求解域的划分,自适应地映射到不同个数的计算核上,从而可利用PC、集群、超算等不同规模的计算资源开展全堆并行模拟。使用针对快堆模拟修改后的子通道模拟软件CTF进行验证,证明了建模方法对于快堆子通道模拟是有效的。基于本文方法在曙光先进计算服务平台上使用两种不同网格规模的算例进行了测试,两组测试最低并行效率在33.02%以上,证明了本文方法的有效性和可用性。     

实验快堆停堆后衰变热特性

一引言无论反应堆是计划内停堆,或是事故工况下的紧急停堆,正确估算停堆后裂变产物的衰变热,对冷却剂丧失事故的安全分析、热量导出系统的合理设计、燃烧过的燃料组件的运输和冷却,以及对全面掌握实验快堆的特性,都有重要的参考价值。计算停堆后的衰变热,一般有两种途径。一种是用停堆后的衰变热积分实验曲线,进行指数多项式符合,然后用符合公式进行计算,这种方法有一定的局限性。另一种是累计法,此法单独处理堆中数百种裂变产物中的每一种裂变产物的衰变热,然后相加求得反应堆总的衰变热。累计法计算的正确性主要依赖于裂变产物数据的正确性,这些数据包括裂变产物产额、半寿命、分支比、衰变方式、发射β     

Numerical study on structural integrity of inner barrel caused by thermal stratification in upper plenum of Monju

In Japanese prototype fast reactor, Monju, an inner barrel with several flow holes is placed at an upper plenum adjacent to a core outlet. When the reactor scram occurs, a cold coolant flows into the bottom of the upper plenum through the core outlet and thermal stratification will appear at the upper plenum. And thus, the inner barrel may be damaged by a thermal stress due to thermal stratification. In this study, a structural integrity assessment method is developed based on fluid-structure interaction analysis and cumulative damage rule. First, a three-dimensional thermal-hydraulics analysis is conducted to simulate a turbine trip test from 40% power operation. Full power output conditions are also simulated by modifying conditions of 40% power output conditions. Next, the thermal stress analysis is modified by adding a practical condition, such as a bending stress. Then, the thermal stress is calculated at each location of the inner barrel. Finally, cumulative damage is evaluated by using the present method. It is concluded that a main factor of cumulative damage is a stress near flow holes that causes stress concentration. It is also found that thermal transient within several hundred seconds after the reactor scram is an important factor.     

Validation of a Computational Simulation Method for Evaluating Thermal Stratification in the Reactor Vessel Upper Plenum of Fast Reactors

Validation of a numerical simulation method is carried out for thermal stratification phenomena in the reactor vessel upper plenum of advanced sodium-cooled fast reactors. The study mainly focuses on the fundamental applicability of commercial computational fluid dynamics (CFD) codes as well as an inhouse code to the evaluation of thermal stratification behavior including the simulation methods such as spatial mesh distribution and RANS-type turbulence models in the analyses. Two kinds of thermal stratification tests are used in the validation, which is done for relatively simple- and conventional-type upper plenum geometries with water and sodium as working fluids. Quantitative comparison between the simulation and test results clarifies that when used with a high-order discretization scheme of the convection term, the investigated CFD codes are applicable to evaluations of the basic behaviors of thermal stratification and especially the vertical temperature gradient of the stratification interface, which is important from the viewpoint of structural integrity. No remarkable difference is seen in the simulation results obtained using different RANS turbulence models, namely, the standard kε model, the RNG k-ε model, and the Reynolds stress model. It is further confirmed in a numerical experiment that the distribution of two or more meshes within the stratification interface will lead to accurate simulation of the interface temperature gradient with less than 10% error.     

压水堆上腔室冷却剂温度振荡现象研究

为了解压水堆上腔室冷却剂的温度振荡现象,利用专业计算软件CFX,采用大涡模拟（LES）方法,对简化上腔室内的瞬态流场进行数值模拟,并与实际数据进行对比分析。结果表明,LES模型可较好地模拟上腔室的温度振荡现象;上腔室出口区域的温度波动多集中于低频部分,未呈现出明显的周期性;出口位置对流场温度分布有明显影响,自入口至上腔室出口中心线所在平面,外围及中心部分温度波动幅度较大,其余区域温度变化幅度较小;而自上腔室出口中心线所在平面至顶部区域,温度波动逐渐趋缓。     

Measurement and analysis of feedback reactivity in the Monju restart core

The purpose of this study is to develop a feedback reactivity measurement technique in the Japanese prototype fast breeder reactor Monju and to validate calculation methodology to forecast the nuclear feedback phenomena. A feedback reactivity measurement technique has been developed based on a reactivity model featuring components that depend on the reactivity coefficients, denoted as reactor power (K_R) and reactor vessel inlet temperature (K_in). This technique can precisely measure the two reactivity components simultaneously and was applied to the feedback reactivity experiment conducted in the Monju system start-up test in May 2010. A thorough evaluation considering all possible biases and uncertainties demonstrated that the reactivity coefficients can be evaluated with a measurement uncertainty smaller than 3%. The evaluated reactivity coefficients were simulated considering the temperature distribution in the core. The calculated and measured values of K_R agreed within 1%, and the value of K_in was consistent with that reported in a previous isothermal temperature coefficient experiment. The measured and calculated fuel subassembly outlet temperatures also agreed well within 0.2°C, which supports the validity of the temperature calculation.     

快堆组件稠密棒束数值模拟

为详细研究快堆组件稠密棒束中的冷却剂流动方式,本工作采用Fluent程序对169棒束快堆燃料组件进行了三维数值模拟,并与已公开发表的文献结果进行了对比。由计算结果可知：计算得到的摩擦系数结果在Re为35885～61354时与试验结果符合较好;从中心到外围,横向流和轴向流在不同的方向和位置呈现出不同的流动特性。根据模拟结果可更准确地预测棒束通道内的流动情况,可为今后稠密棒束组件水力学设计和子通道内流量测量试验提供参考。     

Performance evaluation of eddy current flowmeter in Monju

Measurement of the temperature and flow rate at each fuel subassembly outlet is an effective way for a liquid metal fast breeder reactor to detect a loss of coolant accident or reactivity-initiated accident in the early stage and to understand the reactor’s thermal hydrodynamic performance. Japan Atomic Energy Agency has developed the eddy current flowmeter in practical use and installed 34 of them in the upper core structure of fast breeder reactor, Monju. This report presents data obtained by using the flowmeters in Monju. We observed high linearity between each of the flowmeter’s signal intensity and the primary sodium’s flow rate under 10–100% flow rate condition. High linearity was also observed in a region of low velocity (approx. 0.25 m/s). The fluctuation of flow rate observed by the flowmeters was below 0.2 m/s which is 5% of the time-averaged velocity under a rated condition. These experimental results show that the eddy current flowmeter is an effective tool to detect the changes in relative flow rate.     

高温气冷堆下联箱热气混合的数值模拟和几何优化

高温气冷堆下联箱用于混合温度不均的堆芯出口气体。已有研究显示,当前下联箱设计方案的气体混合能力尚待提高,由其导致的压力损失则需进一步降低。本文采用数值模拟方法,并对比实验数据,讨论了适用于下联箱几何优化的网格类型和网格规模,通过合并原有的肋片状流道以及扩展热气导管起始端,对下联箱的几何形状进行了优化,并通过对比不同优化方案选出了兼具提升气体混合率和减小压力损失的改进方案。本文的研究结果可为高温气冷堆下联箱的改进设计提供参考依据。