快堆主容器内氩气空间传热实验原理及分析

本文研究了池式快堆主容器内氩气空间模拟实验的模拟准则及原理.根据模拟准则和对流换热方程组,研究了模拟实验装置需满足的物理相似和边界相似的条件,并对氩气空间的换热状况进行了计算,确定了实验装置的设计参数范围,从理论上解决了池式实验快堆钠池覆盖气体--氩气空间模拟实验装置的模拟问题.     

CFR600泵支承氩气空间部分三维传热数值模拟

利用流体力学软件Fluent对中国示范快堆（CFR600）氩气空间及相关结构部件进行稳态共轭传热计算,验证分析泵支承结构位于氩气空间部分的热工设计。模拟结果显示,原型设计方案泵支承结构位于氩气空间部分周向温度分布有较大温度梯度,泵支承结构靠近主容器支承径侧温度明显高于远离支承径侧,改进的结构设计减小了泵支承周向的温度梯度,为优化泵支承结构设计提供了技术支撑。     

基于一体化三维数值模拟的中国实验快堆冷钠池及其堆内构件热工特性分析

建立中国实验快堆(CEFR)池式堆本体全尺寸三维模型,进行堆本体冷钠池、热钠池、主容器冷却系统等主要部件的一体化、三维数值计算。通过对热钠池进行部分简化,为冷钠池计算提供更接近实堆运行工况的边界条件,获得CEFR在额定功率稳态工况下冷钠池及其堆内构件三维热工参数,为其结构应力评定及部件设计提供关键输入。计算结果表明:冷钠池内液钠的流动较为复杂,上冷池内流动较为明显;由于冷池中板的阻隔作用,下冷池流动较为微弱。此外,冷钠池内会出现较为明显的热分层现象,使得冷钠池内竖向支承肋板及其堆内构件沿高度方向产生约30℃温差,对其结构强度设计提出更高的要求;主容器冷却系统出口被加热的液钠对上冷钠池的温度、流动分布也有一定影响。本研究为钠冷池式快堆事故安全分析、关键堆内构件结构应力评定及设计提供重要热工输入参数。     

中国实验快堆堆容器冷却系统全厂断电工况温度场分析

堆容器冷却系统是中国实验快堆(CEFR)-回路系统中的重要辅助系统之一,用于在各种工况下对反应堆堆容器进行冷却.本文利用国际通用的计算流体力学软件STAR-CD对CEFR堆容器冷却系统进行三维数值模拟,得到了在全厂断电事故发展过程中堆容器冷却系统的温度场和流场的瞬态分析结果,为相应部件的力学分析以及其它工况的分析提供了数据,对快堆优化设计和安全分析提供了重要的理论支持.     

中国实验快堆燃料破损覆盖气体监测系统工况分析

中国实验快堆燃料破损覆盖气体监测系统投入运行状态时，从反应堆主容器覆盖气体气腔来的高温氩气中含有钠蒸汽，为避免由于温度降低钠蒸汽冷凝成固态堵塞管道，需通过热工计算了解介质从主容器流出进入本系统管道后的传热情况，以采取相应的措施避免上述问题的发生，使系统正常运行。     

示范快堆堆容器及堆内构件的研发设计进展

<正>本工作为国家科技重大专项资助项目。根据快堆三步走战略,在中国实验快堆(CEFR)建造完成后,要进行60万千瓦工业规模的示范快堆电站CFR600的研发,以进行工业规模的示范。示范快堆堆容器及堆内构件是国内首套自主化研发、设计的快堆大型设备,堆内构件涉及内容多,与其相连的相关系统有16个,堆容器上安装的设备有46台套,需实现主热冷却、辅助     

CEFR主容器内正弦三波激励下液面晃动响应

开发了一套可用于估算正弦三波激励下液面晃动对容器壁和顶盖冲击压力的工程方法,计算结果为中国实验快堆(CEFR)主容器及堆内构件的应力分析提供了重要的载荷输入。     

快中子堆主壳顶盖环形空间自然对流的实验研究与数值模拟

快堆顶盖存在一些大小不同的环形空间,这些空间是由穿过顶盖结构的一次泵及热交换器构成的,并使环形空间内的氩气产生自然对流。用有机玻璃建造的实验台,模拟快中子堆主壳顶盖竖直环形空间,进行自然对流实验研究。同时,采用LVEL零方程湍流模型对自然对流进行了数值模拟,所获结果与实验数据基本吻合,表明该模型比常用的kε模型所得出的流型与文献中的和该实验的观测结果更加相符,也能够更加准确地预测环形空间内的温度场与速度场。     

对中国实验快堆旋塞环形空间自然对流的数值模拟

快中子增殖堆中有很多下端开口的环形空间,这些环形空间中氩气的自然对流使得环形空间的内外壁面产生非均匀的热应力,是快堆设计必须解决的问题.针对我国正在进行的快堆的设计要求,对我国第一座快堆(CEFR)环形空间的自然对流进行了数值模拟.模拟采用了LVEL紊流模型.计算结果可供CEFR设计参考.     

CEFR反应堆容器散热计算

CEFR高温的冷却剂与外界大气之间存在巨大的温差。尽管反应堆表面设有保温层,散热量仍很可观。为了确定热量排出情况,保证反应堆安全,本工作对堆容器的散热情况进行计算。计算范围包括主容器、氩气层、保护容器、保温层等。根据边界条件的不同,计算了4种状态下的系统散热情况:额定功率运行、冷停堆、热停堆及全厂断电事故状态。计算考虑热传导、对流及辐射等多种热传递方式。采用热工流体程序STAR-CD,按照1∶1的比例,六面体Hexa网格模拟堆容器结构。计算方法为压力隐式算子分割算法。计算结果显示:堆容器系统的温度由内到外逐渐降低,在…     

示范快堆堆坑通风冷却三维数值模拟

为详细研究示范快堆堆坑内空气流动状态和温度分布情况,检验现行堆坑通风系统布置合理性与冷却效果,本文利用CFD软件对正常运行工况下的示范快堆堆坑空气流域进行三维数值模拟。结果表明,通风系统冷却效果满足设计要求,堆坑混凝土内壁最高温度为50.7 ℃,但堆坑内部流场复杂,温度分布的不均匀性较高,通风系统进出口排布方式需进一步优化。计算结果为主容器及贯穿件支承热工计算提供了更为准确的边界条件,为示范快堆一回路设计提供参考。     

池式钠冷快堆事故源项计算方法研究及其应用

针对传统轻水堆事故源项计算方法不适用池式钠冷快堆的问题,分析可能发生的设计基准事故和超设计基准事故的释放路径,研究建立适用于池式钠冷快堆的堆芯损伤类、泄漏类和钠火类事故源项计算方法。结合示范快堆的6种典型事故：1盒燃料组件瞬时全部堵塞事故、反应堆堆本体覆盖气体边界泄漏事故、一次氩气衰变罐破损事故、主容器泄漏事故、一回路外无保护套管的钠净化管道泄漏事故和一回路无保护套管的外辅助管断裂或泄漏合并隔离阀关不住事故,开展事故源项计算及其剂量后果评价。结果表明：6种事故的放射性后果均低于GB 6249-2011的要求。该方法还可为回路式钠冷快堆、铅铋快堆以及气冷快堆事故源项计算提供参考。     

Study on Calculation Method of Accidental Source Term for Pool-type Sodium-cooled Fast Reactor and Its Application

To deal with the problem that the traditional light water reactor accidental source term calculation method is not suitable for sodium-cooled fast reactor, calculation methods for accidental source term of pool-type sodium-cooled fast reactor, including core damage type, leak type and sodium fire type, were studied and derived on basis of the analysis of release path of potential design basis accidents and beyond design basis accidents. The methods were applied to six typical accidents of the demonstration fast reactor, including the total instantaneous blockage of one fuel assembly, the leakage of cover gas region of reactor main vessel, the damage of primary argon decay tank, the leakage of main vessel, the leakage of sodium purification pipeline without protective sleeve outside the primary circuit, and the leakage of external auxiliary pipeline without protective sleeve outside the primary circuit or the isolation valve tube not be closed. The calculation of accidental source terms and their radiological consequences were carried out. The results show that the radioactive dose consequences of the six accidents are lower than the requirements of GB 6249-2011. The methods proposed can provide reference to the calculations of accidental source term of loop-type sodium-cooled fast reactor, lead-cooled fast reactor and gas-cooled fast reactor.     

Safety approach of DFBR design study in Japan

The goal of the safety design for the demonstration fast breeder reactor is to ensure that the safety level is equivalent to or higher than that of the light water reactors of the same period. The design of the safety features such as reactor shutdown, decay heat removal and confinement systems is of importance to reach the goal. The reactor core is equipped with two independent fast shutdown systems, the primary system and the backup system. In addition, it is planned to strengthen the passive shutdown capability by using self- actuated systems such as a Curie point device for the backup system. The decay heat is removed from the core to the atmosphere through the safety lines of the direct reactor auxiliary cooling system which is composed of four independent lines. Furthermore, under the severe conditions that no active function of the decay heat removal system is available, the heat can be removed by natural convection through the safety lines by taking advantage of the high boiling temperature of sodium. For the confinement function, the reactor vessel is surrounded by a containment vessel and a confinement area.

The design concept of these safety features is described in this paper.     

反应堆压力容器直接安注热分布特性研究

我国非能动系列压水堆将应急冷却系统冷却水的注入管道直接连接于压力容器上,与传统的冷管段安注不同,这种安注方式被称之为反应堆压力容器直接安注。本文以安注条件下的反应堆压力容器为研究对象,采用物理实验与数值分析结合的方法,对安注流体在压力容器表面形成的热分布形态进行研究。研究发现,不同于传统的主管道冷段斜接管安注方式,直接安注条件下安注流体在下降环腔中的分布形态接近于等腰三角形。以实验结果为基础,结合数值计算验证,发现了压力容器热分布角与流速比成正比关系,并进一步提出了安注流体分布计算模型,从而为反应堆安全设计提供参考。      

Design studies on three-dimensional issues for liquid blanket systems in helical reactor FFHR

A new design activity is under way for a helical type DEMO reactor FFHR-d1. The first stage of the activity involves the fundamental issues related to three-dimensional blanket design: (1) the minimum blanket space required for reactor parameter decisions, (2) the support method for the helical blanket system, and (3) the blanket module design. Investigations have been performed with neutronics and mechanical finite-element method calculations. Neutronics investigations indicate that a tungsten carbide radiation shield could reduce the minimum blanket space requirement by ～30 cm at the inboard region of FFHR-d1 compared with the blanket space of ～100 cm in the previous FFHR2 design. The investigations also showed that main shielding materials, ferritic steel and B₄C, could be used separately in a two-layered shielding configuration. The ferritic steel layer of the radiation shield is considered suitable to support the helical blanket system instead of relying on a thin vacuum vessel of the helical reactor. A size of a blanket module for a replacement process and the preferable cooling channel direction under a magnetic field are also discussed.     

用于海上钻井平台的小型钠冷快堆核电源概念设计方案

根据海上石油钻井平台用户电力需求的特点,介绍了一种基于斯特林热气机发电技术的小型钠冷快堆核电源设计方案,研究了小型钠冷快堆核电源的总体技术方案、主回路冷却系统以及关键设备设计方案,并给出小型钠冷快堆核电源的初步布置方案。研究结果表明：小型钠冷快堆核电源概念设计方案符合海上石油钻井平台用户需求的长周期换料、空间限制等特点。     

Thermal-fluid assessment of the design options for reactor vessel cooling in a prismatic core VHTR

The design of the reactor pressure vessel is an important issue in the VHTR design due to its high operating temperature. The extensive experience base in Light Water Reactor makes SA508/533 steel emerge as a strong candidate for the VHTR reactor vessel but requires maintaining the vessel temperature below the ASME code limit. To meet the temperature requirement, three types of vessel cooling options for a prismatic core VHTR are considered: an internal vessel cooling, an external vessel cooling, and an internal insulation. The performances of the vessel cooling options are evaluated by using a system thermo-fluid analysis code and a commercial computational fluid dynamics code during normal operation and accidents. The results suggested that the internal vessel cooling with the modified inlet flow path will be a promising option. The external cooling option does not ensure an effective cooling of the RPV. The insulation option provides an effective reduction of the RPV temperature in the normal and accident conditions but reduces the fuel safety margin during the accidents, requiring careful consideration before the implementation.