CEFR整体冷钠池及其辅助系统温度场三维数值模拟

利用流体力学软件CFX对中国实验快堆（CEFR）整体冷钠池及其辅助系统进行详细稳态模拟计算,并验证运行工况下的热工设计。计算结果表明：冷钠池内存在热分层和热分区现象,确实存在冷钠池上板的高温区,同时泵腔室上部的温度较其他区域的高。本工作为反应堆功率提升及运行提供了技术准备。     

基于附加源项法的钠冷快堆冷热池三维分析

选择计算流体动力学(CFD)为模拟手段,建立快堆一回路钠池的三维闭式一体化CFD模型,对一回路中主要部件进行模拟,其中中间热交换器、独立热交换器、堆芯、主泵采用附加源项法进行模拟,得到中国实验快堆(CEFR)额定功率稳态运行时整个流场的三维速度场与温度场。计算值同CEFR设计值进行比较,结果符合预期,证明了模型的合理性。计算结果表明,钠池较明显地分为温度较低的冷钠池和温度较高的热钠池2个部分,热钠池温差较大,冷热流体搅混现象明显;同时冷钠池、热钠池不同高度的平均温度都很接近,说明分隔冷热钠池的热屏蔽效果较好。     

CEFR3层水平热屏蔽热工流体的数值研究

中国实验快堆（CEFR）采用池式结构,钠池分为冷钠池和热钠池,冷、热钠池之间的3层水平热屏蔽层既是热池的冷边界,也是冷池的热边界,其内部的热工流体直接影响着冷池和热池内的温度分布。针对CEFR满功率运行工况,利用CFX软件,对1/4区域水平热屏蔽层内的热工流体进行三维数值模拟,冷池边界为360℃,热池边界取490和516℃两种方案,数学模型采用RNGk-ε模型。与现有结果相比,由于考虑了对流换热,所得到的平均温度低于设计温度。     

CEFR在ULOF工况下的自然循环能力分析

自然循环能力是衡量钠冷快堆固有安全性的重要指标,堆芯布置、回路设计及工况参数等都会影响堆芯自然循环能力,因此不同堆型的自然循环能力有很大差异。为了保证堆芯事故得到有效缓解,中国实验快堆(CEFR)的设计中通过优化系统布置,重点考虑了堆芯自然循环。本文采用SAS4A程序对CEFR进行系统建模,分析了CEFR在无保护失流(ULOF)工况下的堆芯热工水力参数瞬态特性,验证了CEFR利用自身自然循环和负反馈设计进行事故缓解的能力,本文还对一回路流动阻力和二回路钠装量对堆芯自然循环的影响进行分析。计算结果表明,CEFR具有良好的自然循环特性,在ULOF工况下可以依靠其负反馈停堆,并能够建立起稳定的自然循环从而导出堆芯余热。     

基于SAC-CFR系统分析软件的EBR-Ⅱ LOHSWS事故分析

利用自主开发的系统分析软件SAC-CFR对美国实验增殖堆2号(EBR-Ⅱ)的未能紧急停堆的丧失热阱(LOHSWS)事故全厂瞬态行为进行建模分析。SAC-CFR耦合了新开发的三维钠池计算模型,用于分析EBR-Ⅱ钠池内的流型。结果表明,SAC-CFR计算结果与实验数据相符合,SAC-CFR可用于快堆部分事故工况的瞬态计算,同时也证实了EBR-Ⅱ可在LOHSWS事故下依靠固有安全性停堆。     

基于CFD的铅铋快堆上腔室降阶热分层模型开发

在铅铋快堆紧急停堆后，上腔室发生热分层现象对堆内结构完整性和自然循环余热排出能力产生重要影响，需要重点关注。为克服传统热分层分析方法的缺陷，基于计算流体动力学（CFD）程序Fluent得到高精度的全阶快照，通过特征正交基分解（POD）与Galerkin投影结合的方法构建降阶热分层模型。通过与CFD全阶热分层模型对热分层现象进行对比分析，研究结果表明所开发的降阶热分层模型能很好地模拟上腔室温度分布，能快速地开展铅铋快堆事故下的热分层界面特性研究。本文研究对热分层现象产生机理、有效遏制热分层现象产生提供了重要分析工具。     

多孔介质方法在池式快堆系统分析软件SAC-CFR三维钠池计算模型中的应用

为准确分析池式快堆热钠池内的热工水力学特性,在已开发出的用于池式快堆系统分析的钠池三维计算模型的基础上,应用多孔介质方法建立钠池内中间热交换器、主泵、事故热交换器及屏蔽柱模型,完成了含有多孔介质模型和复杂几何边界的钠池三维计算模型开发。将该模型嵌入池式快堆系统分析软件SAC-CFR后,分析了中国实验快堆在稳态运行和紧急停堆工况下钠池内的流场分布,初步证明了所采用的多孔介质模型的合理性,为下一步非能动余热排出系统模型的开发做准备。     

中国实验快堆额定工况下冷热钠池数值分析

应用计算软件STAR-CD对中国实验快堆(CEFR)正常运行工况中的额定工况进行了三维数值分析，使用多孔介质模型对屏蔽柱的影响进行了模拟，给出了冷热钠池的三维温度场和流场，与已有热工设计进行了比较，并着重分析了浮升力在数值模拟计算中的影响，为事故工况下的设备动态分析及相应的设备力学分析提供了数据。研究结果为CEFR的优化设计及事故分析提供了参考数据和技术支持。     

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。     

中国实验快堆2018年度运行年报

正中国实验快堆(CEFR)热功率为65 MW,试验发电功率为20 MW,首炉燃料使用UO2,采用堆本体池式结构和钠-钠-水三回路热传输系统,并首次设立独立的非能动事故余热排出系统。2018年反应堆处于冷停堆运行状态,继续进行大修遗留工作及大修调试工作,完成系统恢复与功能鉴定,完成3次开堆前检查和常规岛热态运行,对开堆强相关项进行处理和验证,实现冷停堆运行工况下的全厂安全稳定运行。     

Natural circulation characteristics of lead-based reactor under long-term decay heat removal

The natural circulation of primary coolant plays an important role in removing the decay heat in Station-Black-out (SBO) accident from reactor core to decay heat removal systems, such as RVACS and PHXS cooling, for lead-based reactor. In order to study the natural circulation characteristics of primary coolant under Reactor Vessel Air Cooling System (RVACS) and primary heat exchangers (PHXs) cooling, which are crucial to the safety of lead-based reactors. A three-dimensional CFD model for the China Lead-based Research Reactor (CLEAR-I) has been built to analyze the thermal-hydraulics characteristics of primary coolant system and the cooling capability of the two systems. The abilities of the two cooling systems with different decay heat powers were discussed as well. The results demonstrated that the decay heat could be removed effectively only relying on either of the two systems. However, RVACS appeared the obvious thermal stratification phenomenon in the cold pool. Besides, with the increase in decay heat power, the natural circulation capacity of primary coolant between the two systems had a significant difference. The PHXs cooling system was stronger than the RVACS, with respect to the mass flow of primary coolant and the average temperature difference between cold pool and hot pool.     

Analysis of Thermal Stratification Behavior in CEFR Hot Plenum

Thermal stratification which occurs in the CEFR hot plenum after reactor trip has been regarded as an interesting thermal-hydraulics phenomena. The cold sodium in the bottom of the hot plenum will delay the formation of natural circulation in the reactor primary loop, which will have the bad effect on the reactor core cooling after accident. In the views of the integrity of the structure, the appearance of the thermal stratification phenomenon will cause the thermal stress of the reactor main vessel and internals.     

Passive cooldown performance of a 65 MW integral reactor

Advanced integral-type pressurized water reactor with a maximum thermal power of 65 MW is under development at the Korea Atomic Energy Research Institute (KAERI). This 65 MW integral reactor incorporates a number of innovative design features. In the case of a transient, the passive residual heat removal system (PRHRS) is designed to cool the reactor coolant system (RCS) from a normal operation condition to a hot shutdown condition by a natural circulation, and the shutdown cooling system (SCS) is designed to cool the primary system from a hot shutdown condition to a refueling condition by a forced circulation. A realistic calculation has been carried out by using the TASS/SMR code and a sensitivity analysis has been performed to evaluate a passive cooldown capability for various system conditions such as natural and forced circulation conditions for the reactor coolant system or the passive residual heat removal system, and number of active PRHRS trains. The reactor coolant system and the passive residual heat removal system adequately remove the core decay heat by a natural circulation and the 65 MW integral reactor can cool the coolant to the SCS entry condition in the primary system for all the possible operational conditions.     

三环路压水堆压力容器上腔室交混矩阵数值研究

使用STAR-CCM+软件对三环路压水堆压力容器上腔室流场进行了大规模、精细化三维数值模拟,并采用组分跟踪方法分别对157个燃料组件出口冷却剂流动进行计算,构造了一个具有3×157个元素的“上腔室交混矩阵”,用该矩阵即可定量、精确地描述冷却剂从堆芯流出后,经上腔室内交混并再分配到各热管道的复杂流动过程。研究发现堆芯流出的冷却剂在压力容器上腔室内的交混是并不充分的,径向上不同位置燃料组件流出的冷却剂会在上腔室同热管道的接口区域存在明显的对应关系,而燃料组件径向功率分布的差异必然导致热管道中冷却剂热分层现象的产生。      

大规模替代化石及其他有限能源的固有安全高温核动力

为充分发挥核能的巨大潜力,使之在21世纪内早日更大规模地替代煤炭和其他化石与水力能源,本文介绍如何通过创造性地改进融盐冷却球床高温堆-回路的热工水力设计,实现"在任何功率下长期自动运行生产高温核能方法".该方法在充满高沸点(融盐)一次载热剂的深池内,采用特别简单的一体化回路布置,使融盐沿水平方向流过位于池底的环状堆芯,...     

HFETR流量反转分析研究

中国高通量工程试验堆（HFETR）在流量反转过程中,堆芯热工参数的变化会影响到反应堆的安全运行。为此本文利用RELAP5/MOD3程序建立了HFETR模型,进行了相关的研究,得出HFETR强迫循环向自然循环转换的最大允许功率为850kW,自然循环向强迫循环过渡的时刻由压力壳上部水温决定。研究结果表明,现运行模式能保证反应堆的运行安全,为以后运行模式的完善提供了支持。     

中国先进研究堆停堆冷却措施及其主要特点

停堆后冷却问题是中国先进研究堆(CARR)重要的安全问题之一.冷却措施的实施对CARR的安全和建设投资有较重要的影响.CARR采用停堆初期的强迫循环及停堆后期全堆芯自然循环相结合的策略实现正常停堆和事故停堆后的堆芯冷却.停堆冷却的过程具体分为主泵大质量惯性飞轮惰转强迫冷却、应急堆芯冷却系统强迫冷却、自然循环功能部件动作实现全堆芯自然循环3个阶段.3个阶段既相互衔接又相互独立,每个阶段各有特点.停堆冷却策略的实施证明,CARR停堆冷却过程是可靠、有效、合理的,符合先进研究堆的发展趋势.     

Experiments at the mixing test facility ROCOM for benchmarking of CFD codes

For the validation of computational fluid dynamics (CFD) codes, experimental data on fluid flow parameters with high resolution in time and space are needed.Rossendorf Coolant Mixing Model (ROCOM) is a test facility for the investigation of coolant mixing in the primary circuit of pressurized water reactors. This facility reproduces the primary circuit of a German KONVOI-type reactor. All important details of the reactor pressure vessel are modelled at a linear scale of 1:5. The facility is characterized by flexible possibilities of operation in a wide variety of flow regimes and boundary conditions. The flow path of the coolant from the cold legs through the downcomer until the inlet into the core is equipped with high-resolution detectors, in particular, wire mesh sensors in the downcomer of the vessel with a mesh of 64 × 32 measurement positions and in the core inlet plane with one measurement position for the entry into each fuel assembly, to enable high-level CFD code validation. Two different types of experiments at the ROCOM test facility have been proposed for this purpose. The first proposal concerns the transport of a slug of hot, under-borated condensate, which has formed in the cold leg after a small break LOCA, towards the reactor core under natural circulation. The propagation of the emergency core cooling water in the test facility under natural circulation or even stagnant flow conditions should be investigated in the second type of experiment. The measured data can contribute significantly to the validation of CFD codes for complex mixing processes with high relevance for nuclear safety.     

Boron dilution transients during natural circulation flow in PWR—Experiments and CFD simulations

Coolant mixing in the cold leg, downcomer and the lower plenum of pressurized water reactors is an important phenomenon mitigating the reactivity insertion into the core. Therefore, mixing of the de-borated slugs with the ambient coolant in the reactor pressure vessel was investigated at the four loops 1:5 scaled Rossendorf coolant mixing model (ROCOM) mixing test facility. In particular thermal hydraulics analyses have shown, that weakly borated condensate can accumulate in the pump loop seal of those loops, which do not receive a safety injection. After refilling of the primary circuit, natural circulation in the stagnant loops can re-establish simultaneously and the de-borated slugs are shifted towards the reactor pressure vessel (RPV).In the ROCOM experiments, the length of the flow ramp and the initial density difference between the slugs and the ambient coolant was varied. From the test matrix experiments with 0 resp. 2% density difference between the de-borated slugs and the ambient coolant were used to validate the CFD software ANSYS CFX. To model the effects of turbulence on the mean flow a higher order Reynolds stress turbulence model was employed and a mesh consisting of 6.4 million hybrid elements was utilized. Only the experiments and CFD calculations with modeled density differences show stratification in the downcomer. Depending on the degree of density differences the less dense slugs flow around the core barrel at the top of the downcomer. At the opposite side, the lower borated coolant is entrained by the colder safety injection water and transported to the core. The validation proves that ANSYS CFX is able to simulate appropriately the flow field and mixing effects of coolant with different densities.