超临界压力水冷包层方案第一壁的热与应力分析

针对超临界水冷包层中第一壁的运行工况,利用数值计算软件ANSYS中CFX和Workbench两个模块对第一壁结构中的固体域和流体域进行数值分析研究。对比矩形管道和圆形管道内传热及热应力分布发现,矩形管道四个角域强化了壁面流体和主流流体的动量和热量的交换,使传热性能优于圆形管道,而四个角域的存在也造成了该处的应力集中,使结构材料的最大应力明显高于圆形管道。进一步研究冷却剂流向和冷却管道几何结构参数对第一壁结构温度场和应力场的影响发现,在ITER运行工况下,冷却剂流向影响很小,增大冷却管道直径和减小冷却管道最小壁厚均能改善第一壁结构材料中的最高温度,而这两个几何结构参数对第一壁应力的影响较为复杂。     

CFETR氦冷固态包层及其辅助系统动态氚输运分析

实现氚自持、建立完整的氚循环系统并保证氚安全是中国聚变工程实验堆(CFETR)的主要目标之一。在CFETR氦冷固态包层及其辅助系统设计过程中,需对系统级氚输运行为进行详细分析,包括氚滞留量、释放量、浓度的动态变化等。基于已建立的动态氚分析程序TriSim-Dynamic,在此基础上进行修改完善,利用该程序对CFETR氦冷固态包层及其辅助系统氚动态输运进行分析模拟,得到了冷却剂及提氚吹扫气中氚浓度、氚分压,管壁及结构材料中氚盘存量,氚通过包层结构材料和辅助系统管壁向真空室、水冷系统及建筑的渗透通量动态变化,并将其稳态值与已进行基准校核的稳态氚分析程序TriSim-SA及理论解析解进行比较,以初步验证分析结果的准确性,数据结果也对CFETR氚安全分析提供了一定的参考。     

CFETR氦冷陶瓷增殖包层在等离子体主破裂时的电磁结构耦合分析

增殖包层作为中国聚变工程实验堆(China Fusion Engineering Test Reactor,CFETR)的核心部件,承载着能量转换和氚增殖的重要作用。中国科学院等离子体物理研究所在之前增殖包层设计的基础上,又提出了氦冷陶瓷增殖(Helium Cooled Ceramic Breeder,HCCB)包层的概念设计。为评估电磁载荷对HCCB包层结构安全性的影响,借助通用有限元软件ANSYS,研究计算了在等离子体主破裂时包层中产生的感应涡流、洛伦兹力和力矩。通过多物理场耦合分析方法,获取了包层中产生的等效应力和形变位移。结果表明,在等离子体电流指数衰减时,HCCB包层模型上产生的最大等效应力和形变位移满足包层结构设计的要求,同时模拟分析结果也为未来的包层结构优化以及支撑结构设计提供了必要的数据支撑。     

Preliminary thermal hydraulic safety analysis of water-cooled ceramic breeder blanket for CFETR

The Chinese fusion engineering test reactor (CFETR) was expected to bridge from the international thermonuclear experimental reactor (ITER) to the demonstration fusion reactor (DEMO). The water-cooled ceramic breeder (WCCB) blanket is one of the blanket candidates for CFETR. In this paper, preliminary thermal hydraulic safety analyses have been carried out using the system safety analysis code RELAP5 originally developed for light water fission reactors. The pulse operation and three typical loss of coolant accidents (LOCAs), namely, in-vessel LOCA, in-box LOCA, and ex-vessel LOCA, were simulated based on steady-state initialization. Simulation results show that important thermal hydraulic parameters, such as pressure and temperature can meet the design criterion which preliminarily verifies the feasibility of the WCCB blanket from the safety point of view.     

DEMO水冷包层第一壁结构优化设计研究

针对聚变示范堆（DEMO）水冷包层,通过计算流体力学程序CFX和计算结构力学程序ANSYSWorkbench中的SIMULATION模块进行单向流固耦合分析。在对现有设计的DEMO水冷包层第一壁温度和应力数值模拟分析的基础上,改变了第一壁流道结构,着重研究了不同流道结构下的温度和应力分布,分析了几何结构对最高温度和最大应力的影响,提出第一壁结构的优化设计方案。数值模拟结果表明,优化设计方案能有效降低第一壁结构中的最高温度和最大应力。     

CLAM用作超临界水冷包层第一壁结构材料的热与应力性能分析

在聚变堆超临界水冷固态增殖包层第一壁的运行工况下,采用数值方法对采用中国低活化马氏体钢(CLAM)作为结构材料的第一壁进行单向流固耦合分析,为超临界水冷实验包层模块(TBM)的热工设计提供借鉴。分别采用CLAM和F82H作为第一壁结构材料,对比温度场和应力场,并考察不同冷却管道形状(矩形和圆形)、不同冷却管道直径和最小壁厚对第一壁温度场和应力场的影响。结果表明:CLAM的最高温度及最大应力均高于F82H的;采用CLAM作为结构材料时,矩形冷却管道的角域的换热得到了增强,但同时也造成了应力集中,第一壁设计时应综合权衡;增大冷却管道直径和减小最小壁厚均有利于换热。     

Design development and analytical assessment of LLCB TBM in Russian Federation during 2012–2013

The activity on the design, analysis, and R&D for the test blanket module (TBM) with lead–lithium (LL) eutectic coolant and ceramic breeder (CB) was performed in the Russian Federation (RF) according to the technical program of cooperation between the leading research institutes of India (“leader” of the LLCB TBM concept) and RF (“partner”). During the period of 2012–2013, the joint efforts of the RF and Indian specialists were focused on the development of the TBM's basic design with an optimal set of parameters (in particular for testing on both H-H and H-D operation phases of International Thermonuclear Experimental Reactor (ITER) machine). This article briefly describes the results of the TBM design and analysis that have been obtained by the RF specialists (“NIKIET” and D.V. Efremov Institute) in support of the LLCB concept (both DEMO blanket and TBM itself). The main directions of this activity in RF institutes were as follows:

• –development of the TBM design taking into account the ability to manufacture the TBM elements (load-bearing casing, tritium-breeding zone, and attachment system);
• –thermal analysis (in both stationary and transient approaches) of TBM design options (four variations of helium and eutectic flowing directions);
• –structural analysis of TBM design elements for Inductive I operation mode; and
• –recommendations (based upon the results of comparative analysis) on the reference design to be used on further stages of concept development.

The critical issues and further plans on the development of LLCB TBM and corresponding DEMO blanket in the RF are also presented in this article.     

ITER中国液态锂铅实验包层模块第一壁等离子体注入氚滞留分析

对等离子体注入ITER中国液态锂铅实验包层模块第一壁滞留的氚进行了分析,考虑了第一壁温度梯度、材料表面清洁度、加挂Be瓦及结构材料内缺陷等因素对氚滞留量的影响。分析结果显示,滞留的氚主要存在于中子辐照引起的缺陷内;氚滞留量对第一壁面向等离子体侧的清洁度及加挂Be瓦很敏感;总的氚滞留量约0.58 mg,不会对ITER真空室内氚滞留造成显著影响。     

Overview of Helium Cooled Ceramic Reflector Test Blanket Module development in Korea

Korea plans to install and test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in the ITER, because the HCCR blanket concept is one of options of the DEMO blanket. Currently, many design and R&D activities have been performed to develop the Korean HCCR TBM. An integrated design tool for a fusion breeder blanket has been developed based on nuclear technologies including a safety analysis for obtaining a license for testing in the ITER. A half-scale sub-module mockup of the first wall with the manifold was fabricated, and the manufacturability and thermo-hydraulic performances were evaluated. High heat load and helium cooling test facilities have been constructed. Next, the recent status of TBM material development in Korea was introduced including Reduced Activation Ferritic Martensitic (RAFM) steel, lithium ceramic pebbles and silicon carbide (SiC) coated graphite pebbles. Several fabrication methods of RAFM steel, lithium ceramic pebbles, and silicon carbide coating on graphite pebbles were investigated. Recent design and R&D progress on these areas are introduced here.     

Structural components design and analysis of side wall in a Korean HCCR TBM for ITER

Korea has developed a helium cooled ceramic reflector (HCCR) test blanket module (TBM) consisting of four sub-modules in an ITER. From the draft design of the side wall (SW) according to a thermal-hydraulic analysis, a mechanical analysis was performed considering a design channel pressure of 10 MPa. The SW comprised of sixteen grids with the seventeen partitions for the manifold function satisfied 1.5S_m of the allowable stress (S_m) according to RCC-MR code at the maximum stress region in the SW. In addition, an elastic analysis of the draft design of the back manifold (BM) was carried out, which supported the four sub-modules in the HCCR TBM and has the main inlet/outlet of the He cooling pipe, the measurement pipes, and He purge gas lines from the port cell. The results show that the maximum stress was higher than 1.5S_m, and the BM design has been modified to satisfy the BM function and requirements.     

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.     

Z-Pinch驱动的次临界包层的初步热工分析

Z-Pinch产生高能脉冲中子驱动以压水堆乏燃料或天然铀为燃料的次临界包层,以能源输出为主要目标,是实现能源可持续发展的可行途径.本文通过理论计算,分析了次临界包层在连续中子脉冲作用下的材料温度和输出功率随时间的变化规律.结果表明,通过合理的设计可以使材料温度在可接受的范围内波动,系统输出功率在时间轴上可以得到稳定.     

Design and R&D progress of Korean HCCR TBM

Korea plans to test a Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER. The HCCR TBM adopts a four sub-module concept considering the fabricability and the transfer of irradiated TBM for post irradiation examination. Each sub-module has seven-layer breeding zone, including three neutron multiplier layers packed with beryllium pebbles, three lithium ceramic pebble bed packed tritium breeder layers, and a reflector layer packed with graphite pebbles. Based on this configuration, neutronic and electromagnetic calculations were performed and their results were applied for the conceptual design of HCCR TBM that considers manufacturing feasibility. Also, a design and safety analysis of HCCR Test Blanket System (TBS) was performed using integrated design tools modifying nuclear system codes for helium coolant and tritium behavior evaluation. The Advanced Reduced Activation Alloy (ARAA) is being developed as a structural material. A total of 73 candidate ARAA alloys were designed and their out-of-pile performance was evaluated. The graphite pebbles as the neutron reflector were fabricated by using mechanical machining and grounding method with the surface coated with SiC. The hydrogen permeation characteristics of structural materials were evaluated using the Hydrogen PERmeation (HYPER) facility. The recent design and R&D progress on these areas are addressed in this paper.     

全堆芯热工子通道模型的JFNK全局求解方法研究

反应堆热工子通道模型详细考虑了轴向流动、横向交混,湍流交混等多种耦合因素,是堆芯热工水力分析的关键模型,但是这些因素为子通道数值模拟带来了困难和挑战。为了提高热工子通道模型的计算效率和收敛性,本文基于Jacobian-Free Newton-Krylov（JFNK）全局求解方法（以下简称JFNK方法）,开发了全堆芯热工子通道模型的全局求解框架,并基于现有程序的模型和框架建立了基于物理预处理的残差系统,增强JFNK方法的收敛速率。结果表明,JFNK方法的计算效率是固定点迭代方法的5倍,且JFNK方法的效率优势随着收敛精度的提高会更加明显。因此,对于复杂热工子通道模型,JFNK方法有着不错的潜力和效率优势。     

子通道分析程序LINDEN的开发与初步验证

中国广东核电集团有限公司自主开发的子通道分析程序LINDEN采用基于同位网格有限差分技术的四方程漂移流模型以及面向对象的模块化编程技术。该程序具备分析计算的可靠性、稳定性。通过LINDEN和COBRA-Ⅳ程序分别对大亚湾1#、2#机组稳态工况进行了计算分析。结果表明,LINDEN程序和COBRA-Ⅳ程序的计算结果总体吻合较好,LINDEN程序可适用于大型压水堆的热工水力分析。     

Thermal-hydraulic behavior of the NUR nuclear research reactor during a fast loss of flow transient

Nuclear safety analysis remains of crucial importance for both the design and the operation of nuclear reactors. Safety analysis usually entails the simulation of several selected postulated accidents, which can be divided into two main categories, namely reactivity insertion accident (RIA) and loss of flow accident (LOFA). In this paper, thermal-hydraulic simulations of fast LOFA accident were carried out on the new core configuration of the material test research reactor NUR. For this purpose, the nuclear reactor analysis PARET code was used to determine the reactor performance by calculating the reactor power, the reactivity and the temperatures of different components (fuel, clad and coolant) as a function of time. It was observed that during the transient the maximum clad temperature remained well below the critical temperature limit of 110 °C, and the maximum coolant temperature did not exceed the onset of nucleate boiling point of 120 °C. It is concluded that the reactor can be operated at full power level with sufficient safety margins with regard to such kind of transients.     

聚变增殖堆液态金属Li自冷包层的热工水力分析

完成了托卡马克商用混合堆 TCB(Tokamak Commercial Breeder)Li 自冷包层设计的热工水力分析,讨论了热工水力设计中的一些关键问题。用两维有限元热传导程序 AYER 计算了 TCB 包层的温度分布,用液态金属 MHD(Magnetohydraudynamic)压降公式计算了包层的压降。同时,还分析了包层冷却剂丧失事故 LOCA 的瞬态热工过程。分析表明,正常工况下,包层结构材料最高温度,结构材料与冷却剂界面最高温度,以及包层总压降都满足堆设计要求。在 LOCA 工况下,如果停堆后1小时内包层中的燃料球能够借助重力卸出包层,第一壁和包层是安全的,并且不会受到损伤。