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.     

100 MeV Cyclotron LLRF Control Design and Details for RF Circuit

CEFR堆容器及堆内构件是一体化的池式结构。反应堆容器为双层结构，包括主容器和保护容器，堆容器直径约8m，高12．6m。一回路全部主要设备及部分二回路设备置于其中，结构紧凑而复杂。堆容器及部分堆内构件因体积庞大无法整体运至安装现场，因此，按照可运输的尺寸条件分成多个部分在工厂加工制造后运至现场。     

堆本体抗震试验动力相似关系推导与验证

为确保堆本体抗震试验中流体对流效应、脉冲效应和堆本体结构响应的准确性,需保证重力、流体与固体惯性力、结构弹性力和结构应变的相似性。本文从固体结构的振动方程、不可压牛顿流体的动力学方程、流固交界面的边界条件和环形柱体域内液体线性晃动的动力学公式出发,基于控制方程的量纲分析法,推导了考虑液体晃动效应的堆本体地震响应动力相似关系。基于上述相似关系建立了堆容器堆内构件和堆容器内自由液面流体域的缩尺模型,通过有限体积法分析堆容器堆内构件原型和缩尺模型中液体的晃动固有频率、晃动波高、压力以及液体晃动对堆容器支承裙的倾覆力矩。结果表明本文动力相似关系具有合理性和准确性,可用于堆本体缩尺模型的抗震试验研究。     

中国先进研究堆堆芯容器及堆内构件设计研究

堆芯容器及堆内构件是中国先进研究堆(CARR)中的关键设备之一.经过充分调研和精心设计,解决了CARR堆芯容器及堆内构件结构设计中以下几个方面的技术难点:螺纹连接的防松脱设计,填充体与栅板组件的整体装配设计,控制棒导管的上、下支撑设计和密封结构设计等.目前,堆芯容器及堆内构件在现场已安装完毕,经多项设计试验验证表明,达到了预期设计要求.     

反应堆堆芯容器在泄压冲击波作用下的应力分析

１０ＭＷ高温气冷堆是新一代的模块式高温气冷堆。为了分析其堆芯容器在大破口事故下的安全特性，本文研究了堆芯容器在破口泄压冲击波作用下的动态行为，给出了堆芯容器内外两侧的压差瞬变，以及堆芯容器内的应力瞬变，这些数据可为堆芯容器的安全分析和安全设计提供依据。     

与流体晃动相耦合的快堆主容器地震响应的数值模拟

快堆的主容器内存在着自由表面流体,当发生长周期地震时,该流体的晃动会冲击到容器壁,对反应堆造成威胁.本文采用ADINA软件建立快堆主容器的三维有限元模型,模拟了正弦三波激励下液面晃动对容器壁的冲击现象,得到的冲击压力为容器结构完整性分析提供了载荷,验证了运用ADINA软件对自由表面流动进行分析的可行性以及在处理流固耦合问题上的优越性.     

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…     

介绍几种新设计的核燃料元件运输容器

【英国《国际核工程》1981年2月号报道】最近在柏林召开的帕特雷姆会议,就核燃料元件容器新设计作了详细的研究和讨论。根据国际原子能机构对 B(U)型运输容器规定的要求设计出供新的和辐照过的钠冷快堆燃料使用的运输容器,原则上与轻水堆燃料使用的容器相同。但增殖堆的燃料     

基于相似理论的快堆氩气空间温度场实验研究

本文在实验装置上研究了池式快堆主容器内氩气空间温度分布特性。实验结果表明,气体温度在封闭空间中间大部分区域维持一个相对恒定的温度,这是封闭空间内液面自然对流换热的固有特点。根据实验结果分析,提出了氩气空间高度和主容器冷却系统设计应统筹考虑的设计思想,可用于优化池快堆主容器肩部设计,对大型快堆主容器冷却系统优化设计具有参考意义。     

重水堆核电厂典型严重事故氢气风险分析

核电厂严重事故下的氢气控制一直是核电厂关注的热点问题之一。本文采用重水堆一体化事故分析程序建立了主热传输系统(PHTS)模型、排管容器及端屏蔽系统、堆腔以及安全壳模型。分别选取代表高压熔堆和低压熔堆的全厂断电及出口集管大破口失水事故始发严重事故序列,从堆芯氧化产氢以及系统热工水力行为出发,对重水堆产氢特性及点火器的消氢效果进行了研究。分析表明:严重事故下随着堆芯冷却恶化,排管容器内发生锆水反应而产生氢气,排管容器和堆腔内的水对氢气产生有较长时间的抑止作用,随着排管容器和堆腔内水的逐渐烧干,排管容器蠕变失效,熔融堆芯落入堆腔发生堆芯熔融物与混凝土的相互作用而产生大量氢气。当氢气点火器失效时,安全壳隔间内氢气体积份额持续增加,存在燃爆风险;点火器开启时,隔间中的氢气混合气体在较低浓度下点燃,氢气燃烧模式处于慢速燃烧区。     

超临界水冷堆主容器出口管的设计

采用计算流体力学软件CFX对超临界水冷堆主容器出口管的温度分布进行数值分析。研究提出在出口管与主容器之间设计一隔热套管，以避免出口管与主容器直接连接，从而降低主容器与出口管相连接处的温度梯度，提高反应堆的安全性。研究结果表明：隔热套管的几何参数，如直径和长度，对隔热套管与出口管相连接处的温度梯度有一定影响。     

基于移动粒子法的快堆自由表面流体对容器顶盖冲击现象的数值模拟

快堆的主容器内存在自由表面流体，当发生长周期地震时，该流体的晃动有可能冲击到容器顶盖，对反应堆的安全造成威胁。文章引入移动粒子法——MPS方法来模拟流体的运动。在验证了该粒子法对于容器内自由表面晃动问题的准确性和有效性之后，进一步模拟了正弦三波激励下液面晃动对容器顶盖的冲击现象，得到的冲击压力可为容器结构完整性分析提供载荷。     

The evaluation on the effect of shell thickness reduction for ITER vacuum vessel sector 1 and 6

The structural analyses of vacuum vessel have been performed to investigate the effect of shell thickness reduction on structural integrity. The finite element models of vacuum vessel considering original design and thickness reduction have been developed. The expected maximum thickness reduction possibly caused by forming and bending processes during fabrication was applied to the curved region of the analysis models. The linear elastic and nonlinear limit analyses have been performed. The structural integrity of main vessel including lower port stub extension has been verified in accordance with the requirements of RCC-MR. It is concluded that the inner and outer shells of main vessel still have enough strength margins under pressure and VDE (Vertical Displacement Event) load conditions in spite of thickness reduction. These results have been reviewed and approved by ITER Organization and ANB (Agreed Notified Body).     

Experimental investigation of weir instability in main vessel cooling system of 1/4 FBR model

The 500 MWe Prototype Fast Breeder Reactor (PFBR) is under construction at Kalpakkam, India. The main vessel of this pool type reactor acts as the primary containment in the reactor assembly. In order to keep the main vessel temperature below creep range and to reduce high temperature embrittlement and also to ensure its healthiness for 40 years of reactor life, a small fraction of core flow (0.5 m³/s) is sent through an annular space formed between the main vessel and a cylindrical baffle (primary thermal baffle) to cool the vessel. The sodium after cooling the main vessel overflows the primary baffle (weir shell) and falls into another concentric pool of sodium separated from the cold pool by the secondary thermal baffle and then returned to cold pool. The weir shell, where the overflow of liquid sodium takes place, is a thin shell prone to flow induced vibrations due to instability caused by sloshing and fluid-structure interaction. A similar vibration phenomenon was first observed during the commissioning of Super-Phenix reactor. In order to understand the phenomenon and provide necessary experimental back up to validate the analytical models, weir instability experiments were conducted in a 1:4 scale stainless steel (SS) model installed in a water loop. The experiments were conducted with flow rate and fall height as the varying parameters. The experimental results showed that the instability of the weir shell was caused due to fluid structure interaction. This paper discusses the details of the model, the modeling laws, similitude criteria adopted, analytical prediction, the experimental results and conclusion.     

燃料组件冲刷实验压力容器的设计

本文介绍了秦山核电厂一期工程1:1燃料组件高温高压冲刷实验压力容器的设计。该设备的主要特点是:无级调节错对中量,弹性支撑,可做横向冲刷实验,并可用于600MW,900MW和1200MW核电厂全尺寸燃料组件的冲刷实验。     

The design of the prestressed concrete reactor vessel for gas-cooled heating reactors

The Gas-cooled Heating Reactor (GHR) based on the pebble red reactor principle was developed by ABB/HRB. An essential part of this concept is the prestressed concrete reactor vessel in which the liner cooling system acts as a heat exchanger. As a main design feature the vessel is designed so that failure can be safely ruled out under all operating and accident conditions. It is of great advantage that the liner is not exposed to primary stresses and that corrosion can be excluded because of the environmental conditions. Relevant material flaws are ruled out by the considerably extent and level of quality assurance measures. A special heat-resistant concrete developed by HRB will be used for the prestressed concrete structure. Its strength behaviour is characterized by only a small reduction during normal operation and also under accident conditions. Even in the event of a hypothetical accident the integrity of the vessel remains intact. Thus the GHR offers a simple, safe and economic source of heat generation.     

Corrosion of CEFR Main Vessel Material Weldment in High Temperature Sodium

CEFR main vessel is a first grade nuclear safety unit that cannot be replaced after building of reactor, therefore, the material for main vessel should be carefully concerned. In order to ensure the safety of the reactor, it is necessary to evaluate the various properties of the main vessel material, especially the weldment. The inter-granular corrosion resistance in high temperature sodium is one of them.     

Radiation Service Life of the Vessel Metal in Operating VVER Reactors

The current status of the prediction of radiation embrittlement of the vessel material in first- and second-generation VVER reactors is analyzed. The radiation service life of the vessel of each type of reactor is determined by factors due to the special features of the operating regime of the reactor and the chemical composition of the vessel metal. A method of monitoring the state of the material of first-generation reactor vessels is examined. The method is based on extracting and studying samples of a metal from the inner surface of the sample. The main problems of monitoring the state of the metal in VVER-440/213 and VVER-1000 vessels are analyzed. It is indicated that adjustments must be made in the normative relations which are currently used for predicting radiation embrittlement of vessel material. The most important questions concerning reactor dosimetry for VVER vessel material are illuminated.__________Translated from Atomnaya Energiya, Vol. 98, No. 6, pp. 460–472, June 2005.     

Manufacturing and assembly of the plasma- and outer vessel of the cryostat for Wendelstein 7-X

Wendelstein 7-X is an advanced helical stellarator, which is presently under construction at the Greifswald branch of IPP. A set of 70 superconducting coils arranged in five modules provides a twisted shaped magnetic cage for the plasma and allows steady state operation. Operation of the magnet system at cryogenic temperatures requires a cryostat which provides thermal protection and gives access to the plasma. The main components of the cryostat are the plasma vessel, the outer vessel, the ports, and the thermal insulation. The German company, MAN Diesel & Turbo SE Deggendorf (former MAN DWE GmbH Deggendorf), is responsible for the manufacture and assembly of the plasma vessel, the outer vessel and the thermal insulation. This paper describes the manufacturing and assembly technology of the plasma and outer vessel of the cryostat for Wendelstein 7-X.