中国工程试验堆堆芯入口流量分配特性实验研究

堆芯入口流量分配研究是新型反应堆设计过程中一项重要的工程验证实验,其结果能为反应堆的热工水力及安全分析提供数据支撑。本文针对中国工程试验堆（CENTER）,采用缩比模型开展了堆芯入口流量分配特性实验研究,在不同工况下获得了模拟燃料组件、铍/铝组件、钴靶组件及控制棒导向管内的流量分配因子。实验结果表明：在本文研究的工况范围中,堆芯中大部分冷却剂流过模拟燃料组件,同类型模拟组件间的流量分配较均匀,最大流量相对偏差在±4%以内。实验入口总流量对流量分配特性几乎没有影响。     

秦山核电二期工程反应堆水力模拟实验研究

该实验研究采取了理论计算、单向实验和反应堆整体水力实验相结合的技术路线.反应堆整体实验模型的比例为14.模拟燃料组件按开式栅格模拟原理设计为2×2棒束组件,其轴向和横向流动特性分别与原型相同,每个组件的入口段装有测量流量用的特制涡轮流量计和测量浓度用的微型电导电极.实验回路由额定流量为2×1170m3/h的两对称环路组成.实验得到的堆芯流量分配、反应堆各部分阻力系数、各部位旁漏流量和堆芯入口腔的交混因子等结果数据,验证并优化了反应堆的结构设计,为反应堆热工水力设计和安全分析提供了必需的和可靠的输入参数.     

CAP1400反应堆堆芯入口流量分配试验研究

反应堆堆芯入口流量分配是反应堆水力性能研究的重要内容之一,其与堆芯热裕量和燃料组件燃料棒的流致振动密切相关,从而影响反应堆的运行。CAP1400反应堆堆芯入口流量分配试验是验证CAP1400反应堆结构设计与分析的一个重要环节,旨在验证CAP1400反应堆堆芯入口流量分配的均匀程度。本文通过1/6比例模型试验,获得无均流板结构工况和带均流板结构3种工况（均匀流量工况、非均匀流量工况、偏回路流量工况）下CAP1400反应堆堆芯入口流量分配结果,并进行了各工况下流量分配均匀程度的分析。试验结果表明,CAP1400反应堆堆芯入口具有较好的流量分配效果。     

反应堆控制棒导向管环形缝隙内漏流量的实验测定及其研究

从研究流量测试的特点着手,对流经电功率为30万千瓦的压水堆核电站反应堆控制棒导向管环形缝隙内的漏流量进行了实验测定。为验证该反应堆现有燃料组件在导向管上的开孔尺寸提供了可靠的实验数据。     

铅铋反应堆堆芯流量分区智能优化方法研究

堆芯流量分区是实现堆芯出口温度展平的重要手段,合理地分区可以提高反应堆的安全性和经济性。本文将人工智能优化算法与单通道模型进行耦合,构建了反应堆堆芯流量分区计算模型,分别开展遗传算法、差分进化算法、量子遗传算法在反应堆流量分区问题上的收敛性分析。根据所得最优算法,分别以寿期初功率分布、各燃料组件在整个寿期内最大功率为样本数据,基于小型长寿命自然循环铅铋快堆SPALLER -100开展两种不同流量分区方案对比分析。研究结果表明,在3种智能优化算法中,量子遗传算法在反应堆流量分区问题上收敛性最佳,能较快地搜索到最优分区结果;基于寿期初功率分布样本数据所得燃料组件最大出口温度超出反应堆热工安全限值,而基于各燃料组件在整个寿期内最大功率所得燃料组件最大出口温度降低了140 K,且始终保持在热工安全限值之下;SPALLER-100反应堆最佳分区数为5,再增加分区数对提高反应堆热工安全性能影响较小。      

控制棒导向管外流道旁流特性实验研究

冷却剂流经核反应堆堆芯时,绝大部分通过燃料组件内部流过,带走裂变能量。另外一小部分作为旁流经过燃料组件外侧流道、控制棒导向管外侧及内侧流道流出。为确保反应堆在正常运行工况下的安全性,必须限制堆芯旁流流量。本文通过开展导向管外侧流道阻力特性实验研究,在不同流量工况下获得了分段压差,并进一步拟合了雷诺数与阻力系数的关系式。实验结果表明,导向管外侧流道压力损失主要集中在堆芯下栅格板处,当反应堆额定工况运行时,单组导向管外侧流量仅为0.196 m³/h。     

铅基研究堆燃料组件阻力特性模拟实验与分析

燃料组件是反应堆的核心部件,冷却剂在堆芯组件内部流动的流动阻力特性是反应堆热工水力特征的重要参数之一。本文以中国铅基研究堆(CLEAR-I)燃料组件为实验模型,利用水作为工作介质,基于雷诺数Re相似准则,间接研究燃料组件在铅基合金冷却剂中的阻力特性,通过测量常温水在不同流速下流经燃料棒束产生的压降值,获得Re在4000~43500范围内摩擦因子随Re变化的关系式,并将阻力模型Rehme关系式和Novendstern关系式的理论分析与实验数值进行了对比研究,结果表明,两个阻力计算模型与实验最大相对误差分别为18.9%和35.6%。     

中国工程试验堆堆芯入口流量分配特性实验研究

堆芯入口流量分配研究是新型反应堆设计过程中一项重要的工程验证实验,其结果能为反应堆的热工水力及安全分析提供数据支撑。本文针对中国工程试验堆(CENTER),采用缩比模型开展了堆芯入口流量分配特性实验研究,在不同工况下获得了模拟燃料组件、铍/铝组件、钴靶组件及控制棒导向管内的流量分配因子。实验结果表明:在本文研究的工况范围中,堆芯中大部分冷却剂流过模拟燃料组件,同类型模拟组件间的流量分配较均匀,最大流量相对偏差在±4%以内。实验入口总流量对流量分配特性几乎没有影响。     

中国加速器驱动嬗变研究装置堆芯流量分配设计与优化

针对中国加速器驱动嬗变研究装置的液态铅铋冷却反应堆,采用计算流体力学软件对燃料组件的上下管座段以及堆芯的流动传热进行了三维计算。针对上下管座段的水力学分析,得到了部件阻力系数与流速、开口面积等参数的关系,为堆芯流量分配的设计工作奠定了基础。基于上述结果,采用多孔介质模型建立了全堆流动传热分析模型,针对流量分配问题进行了数值模拟,以功率份额为流量分配的参考依据,通过调整每盒燃料组件入口面积的大小,使得各个组件的流量分配份额与功率份额基本一致,冷却剂在组件出口处的温度分布得到了较好的展平。     

脉动流下棒束通道内相位差及瞬态流场研究

事故条件及海洋条件下反应堆处于非稳态工况,堆芯燃料组件内热工水力行为具有瞬变及多因素耦合特性,对反应堆的安全提出更高挑战,因此有必要对燃料组件内瞬态特性进行研究。本文通过测量棒状燃料组件内压降和流量之间延迟时间开展棒束通道脉动流条件下相位差研究,对比了相位差在不同振幅、不同流动状态下的变化特性,并分析了定位格架对脉动流相位差的作用特点。另外,基于粒子图像测速（PIV）技术开展了脉动流条件下棒束通道内流场分布特性研究,对比了相同流量条件下稳态工况与瞬态工况下流场分布差异,分析了主流具备不同加速度时棒束通道内流场分布特征。实验结果表明：定位格架可减小脉动流下棒束通道内相位差;棒束通道内流场演化滞后于主流量变化。实验结果有助于揭示燃料组件在非稳态条件下瞬态特性,并为燃料组件的设计和优化奠定基础。     

堆芯燃料组件流阻特性模块式与整体式三维数值模拟方案比较

组件的阻力特性影响堆芯不同类型组件的流量分配,对堆芯的设计起到至关重要的影响。为提高验证堆芯燃料组件特性的求解精度及效率,本文针对燃料区6类燃料组件中的两类进行模块式及整体式三维数值模拟,获得了两类组件的流阻特性,并用相同条件下的全组件试验结果进行了对比。结果表明：推广至堆芯所有燃料组件流阻特性预测,模块式所需计算时间约为整体式的1/6,但整体式三维数值模拟所得压降与试验结果吻合度高,误差较模块式小。最后深入研究了流速及温度变化对流阻特性的影响。该研究为后续各类组件的流阻特性研究方法选取提供技术支持。     

大型先进压水堆核电站堆芯装载方案设想

核电站堆芯装载方案是反应堆堆芯设计的重要基础,它首先必须满足核安全的要求,同时还要尽可能地提高经济性。通过分析国内、外百万千瓦级核电站的堆芯装载,对反应堆输出功率、燃料组件数、堆芯平均线功率密度进行比较,给出我国大型先进压水堆核电站示范工程反应堆堆芯装载方案的设想,为技术决策提供参考。     

Thermal and stability considerations for a supercritical water-cooled fast reactor with downward-flow channels during power-raising phase of plant startup

This paper describes study on the procedure of raising the reactor thermal power and the reactor coolant flow rate during the power-raising phase of plant startup for the supercritical water-cooled fast reactor (SWFR), which is selected as one of the Generation IV reactor concepts. Since part of the seed fuel assemblies and all the blanket fuel assemblies of the SWFR are cooled by downward flow, the feedwater from the reactor vessel inlet nozzle to the mixing plenum located below the core is distributed among these fuel assemblies and the downcomer. The flow rate distribution as the function of both the reactor thermal power and the feedwater flow rate, which are the design parameters for the power-raising phase, is obtained by the thermal hydraulic calculations. Based on the flow rate distribution, thermal analyses and thermal-hydraulic stability analyses are carried out in order to obtain the available region of the reactor thermal power and the feedwater flow rate for the power-raising phase. The criteria for the “available” region are the maximum cladding surface temperature (MCST) and the decay ratio of thermal-hydraulic stability in three “hot” channels; two seed assemblies with upward/downward flow and a blanket assembly. The effects of various heat transfer correlations and axial power distributions are also studied.     

Adaptation of Approved Transport and Storage Cask Lines to Individual Customer Requirements

Abstract

Within the decommissioning programmes of the Italian nuclear power plants, the Italian multi-utility company ENEL decided to rely on on-site dry storage while waiting for the availability of the national interim storage site. SOGIN (Società Gestione Impianti Nucleari SpA, Rome, Italy), now in charge of all nuclear power plant (NPP) decommissioning activities was created in the ENEL group but is now owned by the Italian government. In 2000 it ordered 30 CASTOR® casks for the storage of its spent fuel not covered by existing or future reprocessing contracts. Ten CASTOR X/A17 casks will contain the Trino pressurised water reactor (PWR) fuel and the Garigliano boiling water reactor (BWR) fuel currently stored in pools at the nuclear power plant Trino and the Avogadro nuclear facility at Saluggia. Additionally 20 CASTOR X/B52 casks will contain the BWR fuel assemblies, which are stored in the pool at the Caorso nuclear power plant. GNB (Gesellschaft fuer Nuklear-Behaelter mbH, Essen, Germany) has completed detailed studies for the design of both types of cask. The tailored cask design is based on the well-established and proven design features of CASTOR reference casks and is responsive to the needs and requirements of the Italian fuel and handling conditions. The design of the CASTOR X/A17 for up to 17 Trino PWR fuel assemblies or 17 Garigliano BWR fuel assemblies and the CASTOR X/B52 cask holding up to 52 Caorso BWR fuel assemblies is suitable for the following conditions of use: loading of the casks in the fuel pools of the nuclear installations at Trino, Caorso and Avogadro; no upgrading of the Current on-site crane capacities; transport of the fuel assemblies, which are currently stored at the Saluggia facility to the nuclear power plant Trino; on-site storage in a vertical or horizontal position with the possibility of transfer to another temporary storage or a final repository, even after a number of years; the partial loading of mixed oxide (MOX) and failed fuel; loading and drying of bottled Garigliano fuel assemblies. On the basis of the CASTOR V/19 and CASTOR V/52 cask lines, the design of the CASTOR X/A17 and X/B52 casks aims at optimising safety and economics under the given boundary conditions. The long time for which fuel is kept in intermediate wet storage results in a reduced shielding and thermal-conduction requirement. This is used to meet the tight mass and geometry restrictions while allowing for the largest cask capacity possible.     

钠冷快堆分析程序ATHAS-LMR的子通道模型

以子通道模型和绕丝分布式阻力模型为基础,研发了液态金属快中子增殖堆热工水力子通道分析程序ATHAS-LMR,以对液态金属快中子增殖堆燃料组件中的热工水力现象进行分析。与国外知名实验和类似子通道分析程序比较,结果表明：ATHAS-LMR与实验结果及其他子通道分析程序的结果相近,能够完成包括堵流工况的各种工况下液态金属快中子增殖堆组件的热工水力性能分析。     

海上浮动堆燃料组件进入堆芯仿真分析

与陆上核电厂不同,海上浮动堆换料操作会受海浪环境的影响,因此对换料操作工艺和设备提出了新要求。本文选取海洋核动力平台的海上换料方案,对燃料组件在摇摆工况进入堆芯过程进行了仿真分析。分析结果表明,引入万向节的燃料组件进入堆芯过程中,燃料组件满足强度设计要求。      

Control of a High Temperature Supercritical Pressure Light Water Cooled and Moderated Reactor with Water Rods

The plant system of a supercritical pressure light water reactor (SCR) is once-through direct cycle. The whole coolant from the feedwater pumps is driven to the turbines. The core flow rate is less than 1/7 of that of a boiling water reactor. In the present design of the high temperature thermal reactor (SCLWR-H), the fuel assemblies contain many water rods in which the coolant flows downward. The stepwise responses of the SCLWR-H are analyzed against perturbations without a control system. Based on these analyses, a control system of the SCLWR-H is designed. The pressure is controlled by the turbine control valves. The main steam temperature is controlled by the feedwater pumps. The reactor power is controlled by the control rods. The control parameters are optimized by the test calculations to satisfy the criteria of both fast convergence and stability. The reactor is controlled stably with the designed control systems against various perturbations, such as setpoint change of the pressure, the main steam temperature and the core power, decrease in the feedwater temperature, and decrease in the feedwater flow rate.     

商用压水堆核燃料研发进展与应用展望

压水堆(PWR)是目前核电厂反应堆的主力堆型,而核燃料是反应堆的能量源泉和放射性裂变物质的主要来源,关乎核电厂的经济性和安全性。本文对当前国际上面向商用PWR应用研发的掺杂UO₂燃料、高铀密度燃料、微封装燃料和金属燃料的性能特点、技术状态及前景进行了归纳和评价。在掺杂UO₂燃料中,大晶粒燃料具有较高的技术成熟度,将在PWR实现大规模商用;高铀密度燃料和金属燃料在高温水腐蚀氧化问题以及事故下的行为仍待研究解决;具有极致安全的微封装燃料更适合特殊用途的小型反应堆。应协同开展先进燃料组件设计、建立设计准则以及研发高保真的性能分析技术等,以充分发挥新型燃料的可靠性及高燃耗优势。     

AP1000 core design with 50% MOX loading

The European uility requirements (EUR) document states that the next generation European passive plant (EPP) reactor core design shall be optimized for UO₂ fuel assemblies, with provisions made to allow for up to 50% mixed-oxide (MOX) fuel assemblies. The use of MOX in the core design will have significant impacts on key physics parameters and safety analysis assumptions. Furthermore, the MOX fuel rod design must also consider fuel performance criterion important to maintaining the integrity of the fuel rod over its intended lifetime. The purpose of this paper is to demonstrate that the AP1000 is capable of complying with the EUR requirement for MOX utilization without significant changes to the design of the plant. The analyses documented within will compare a 100% UO₂ core design and a mixed MOX/UO₂ core design, discussing relevant results related to reactivity management, power margin and fuel rod performance.