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。     

He冷却LLi氚增殖包层的热工水力初步设计

完成了托卡马克工程试验混合堆ＴＥＴＢ－Ⅲ Ｈｅ冷液态金属Ｌｉ（ＬＬｉ）氚增殖包层的初步热工水力设计，探讨了包层中载氚的两种可能方式，同时，用程序完成了对第一壁和包层的温度场计算及热工水力设计参数的初步优化。分析结果表明，尽管Ｈｅ气的导热性和密度都比液态金属冷却剂低得多，但仍有可能使堆芯在２．０ＭＰａ的低压下运行，并且包层的热工水力设计参数满足设计要求。     

CFETR水冷陶瓷增殖剂包层模块产氚率实验验证

水冷陶瓷增殖剂（WCCB）包层作为中国聚变工程试验堆（CFETR）候选包层之一,承担着氚增殖、核热提取、屏蔽等重要涉核功能,其中子学设计的可靠性直接影响CFETR氚自持目标的实现。为验证中子学设计工具,即MCNP和FNEDL3.0数据库,在WCCB包层中子学设计中的可靠性,基于研制出的WCCB包层模块,在DT中子环境下开展中子学实验,对以产氚率（TPR）为代表的中子学参数进行了模拟值（C）和实验值（E）对比分析。结果表明,模块中轴线位置处TPR的C/E为0.97?1.08,而模块边缘位置处TPR的C/E为0.65?0.82;模块钛酸锂层边缘区197Au（n,γ）198Au反应率的C/E为0.72?0.90,表明模块边缘区存在非期望的散射中子,导致该区TPR模拟值和实验值偏离较大。     

Neutronics Analysis of Water-Cooled Ceramic Breeder Blanket for CFETR

In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5otorus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models,shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1,the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined.The results indicate that the global TBR of no less than 1.2 will be a big challenge for the watercooled ceramic breeder blanket for CFETR.     

CFETR氦冷固态增殖剂包层产氚性能研究及优化

氦冷固态增殖剂包层是中国聚变工程实验堆（CFETR）的3种候选包层概念之一。本文基于中国核工业西南物理研究院提出的一种氦冷固态增殖剂包层概念,通过蒙特卡罗输运程序MCNP5建立了包层三维中子学模型,探究了不同几何布置方案及结构设计参数对包层产氚性能的影响,得到了全堆氚增殖比（TBR）及极向各包层模块产氚分布,并由优化后的模型得到了包层模块核热分布。结果表明,优化后的TBR达到1.177,满足氚自持的最低要求。     

Neutronics Comparison Analysis of the Water Cooled Ceramics Breeding Blanket for CFETR

China Fusion Engineering Test Reactor(CFETR) is an ITER-like fusion engineering test reactor that is intended to fill the scientific and technical gaps between ITER and DEMO.One of the main missions of CFETR is to achieve a tritium breeding ratio that is no less than 1.2to ensure tritium self-sufficiency.A concept design for a water cooled ceramics breeding blanket(WCCB) is presented based on a scheme with the breeder and the multiplier located in separate panels for CFETR.Based on this concept,a one-dimensional(1D) radial built breeding blanket was first designed,and then several three-dimensional models were developed with various neutron source definitions and breeding blanket module arrangements based on the 1D radial build.A set of nuclear analyses have been carried out to compare the differences in neutronics characteristics given by different calculation models,addressing neutron wall loading(NWL),tritium breeding ratio(TBR),fast neutron flux on inboard side and nuclear heating deposition on main in-vessel components.The impact of differences in modeling on the nuclear performance has been analyzed and summarized regarding the WCCB concept design.     

An Analysis of Ripple and Error Fields Induced by a Blanket in the CFETR

The Chinese Fusion Engineering Tokamak Reactor(CFETR) is an important intermediate device between ITER and DEMO. The Water Cooled Ceramic Breeder(WCCB)blanket whose structural material is mainly made of Reduced Activation Ferritic/Martensitic(RAFM) steel, is one of the candidate conceptual blanket design. An analysis of ripple and error field induced by RAFM steel in WCCB is evaluated with the method of static magnetic analysis in the ANSYS code. Significant additional magnetic field is produced by blanket and it leads to an increased ripple field. Maximum ripple along the separatrix line reaches 0.53% which is higher than 0.5% of the acceptable design value. Simultaneously, one blanket module is taken out for heating purpose and the resulting error field is calculated to be seriously against the requirement.     

CFETR氦冷陶瓷增殖包层中子学分析

为满足中国聚变工程实验堆(CFETR)包层的应用要求,本文提出氦冷陶瓷增殖(HCCB)包层方案。为验证HCCB包层设计方案的合理性与可行性,采用三维蒙特卡罗粒子输运程序MCNP,计算和分析了HCCB包层方案的氚增殖比、中子壁负载、中子通量密度、核热、辐照损伤等中子学特性。结果表明,HCCB包层方案满足氚自持要求,中子通量密度和核热分布合理,屏蔽性能良好,基本满足设计要求。     

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.     

Thermal Hydraulic Design and Analysis of a Water-Cooled Ceramic Breeder Blanket with Superheated Steam for CFETR

The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for China fusion engineering test reactor (CFETR). In order to improve power generation efficiency and tritium breeding ratio, WCCB with superheated steam is under development. The thermal-hydraulic design is the key to achieve the purpose of safe heat removal and efficient power generation under normal and partial loading operation conditions. In this paper, the coolant flow scheme was designed and one self-developed analytical program was developed, based on a theoretical heat transfer model and empirical correlations. Employing this program, the design and analysis of related thermal-hydraulic parameters were performed under different fusion power conditions. The results indicated that the superheated steam water-cooled blanket is feasible.     

Analysis of Time-Dependent Tritium Breeding Capability of Water Cooled Ceramic Breeder Blanket for CFETR

Attaining tritium self-sufficiency is an important mission for the Chinese Fusion Engineering Testing Reactor(CFETR) operating on a Deuterium-Tritium(D-T) fuel cycle. It is necessary to study the tritium breeding ratio(TBR) and breeding tritium inventory variation with operation time so as to provide an accurate data for dynamic modeling and analysis of the tritium fuel cycle. A water cooled ceramic breeder(WCCB) blanket is one candidate of blanket concepts for the CFETR. Based on the detailed 3D neutronics model of CFETR with the WCCB blanket,the time-dependent TBR and tritium surplus were evaluated by a coupling calculation of the Monte Carlo N-Particle Transport Code(MCNP) and the fusion activation code FISPACT-2007.The results indicated that the TBR and tritium surplus of the WCCB blanket were a function of operation time and fusion power due to the Li consumption in breeder and material activation.In addition, by comparison with the results calculated by using the 3D neutronics model and employing the transfer factor constant from 1D to 3D, it is noted that 1D analysis leads to an over-estimation for the time-dependent tritium breeding capability when fusion power is larger than 1000 MW.     

中国HCCB-TBM氚增殖单元三维多物理场耦合氚输运模拟分析

中国氦冷固态实验包层模块(HCCB-TBM)将在国际热核聚变实验堆(ITER)上安装测试,以验证其氚增殖能力与核热移出能力。HCCB-TBM中的氚输运与流体的传热和传质、氢同位素交换、结构材料的SORET效应密切相关。考虑以上物理因素,基于商业软件COMSOL完成了HCCB-TBM氚增殖单元多物理场耦合的氢同位素输运模拟分析。分析结果表明:球床吹洗气体中含氢有助于抑制氚渗透损失;当吹洗气体含氢浓度为4.66×10^-2 mol/m³时,产生约13.2倍的氚渗透阻止效应。     

基于RELAP5的CFETR氦冷陶瓷增殖包层模块热工安全分析

本文对中国聚变工程实验堆(CFETR)氦冷陶瓷增殖(HCCB)包层进行热工安全分析。采用大型反应堆瞬态分析程序RELAP5对HCCB包层建模,并进行稳态分析和假设事故的模拟。计算结果表明,CFETR HCCB包层在真空室内氦气泄漏和增殖区盒内氦气泄漏事故中均未出现结构材料熔化,同时各部分的压强变化情况均未超出设计阈值,包层系统在事故发生后均能有效快速地排出余热。CFETR HCCB包层的设计满足热工安全方面的要求。     

CFETR第一壁及赤道面外包层中子辐照损伤初步分析

中国聚变工程实验堆(Chinese Fusion Engineering Testing Reactor,CFETR)的包层和偏滤器第一壁面向堆芯等离子体,第一壁辐照损伤分析对于托克马克安全运行至关重要。赤道面外包层较其它包层距离堆芯等离子体中心更近,其结构材料承受中子辐照大。因此,进行中子辐照损伤评估十分必要。基于此目的,采用计算机辅助设计(Computer Aided Design,CAD)模型和蒙特卡罗中子学建模转换接口Mc CAD完成中子学建模,并用蒙特卡罗方法的粒子输运程序计算第一壁和氦冷固态外包层结构材料辐照损伤。此外,对比了铍和钨作为面向等离子体材料两种情况下第一壁的受损情况。计算结果表明,氦冷固态包层模型下结构材料可以满足CFETR一期的运行要求。     

Concept design on RH maintenance of CFETR Tokamak reactor

CFETR which stands for Chinese Fusion Engineering Testing Reactor is a superconducting Tokamak device. The concept design on RH maintenance of CFETR has been done in the past year. It is known that, the RH maintenance is one of the most important parts for Tokamak reactor. The fusion power was designed as 50–200 MW and its duty cycle time (or burning time) was estimated as 30–50%. The center magnetic field strength on the TF magnet is 5.0 T, the maximum capacity of the volt seconds provided by center solenoid winding will be about 160 VS. The plasma current will be 10 MA and its major radius and minor radius is 5.7 m and 1.6 m respectively. All the components of CFETR which provide their basic functions must be maintained and inspected during the reactor lifetime. Thus, the remote handling (RH) maintenance system should be a key component, which must be detailedly designed during the concept design processing of CFETR, for the operation of reactor. The main design work for RH maintenance in this paper was carried out including the divertor RH system, the blanket RH system and the transfer cask system. What is more, the technical problems encountered in the design process will also be discussed.     

Modeling and multi-pipe manifolds optimization of the WCCB blanket sector for CFETR

In this paper, one standard water cooled ceramic breeder blanket sector has been modeled for the Chinese fusion engineering test reactor using RELAP5/MOD3.3 with details of anisotropic structures, positions and nuclear heat of the blanket modules. The multi-pipe manifolds of the current sector design scheme has been designed and analyzed. And an optimized scheme was proposed to further reduce the pressure drop, uniform the flow distribution, and prevent overheating. Also the fusion power excursion transients were simulated to evaluate the system heat removal and recovery ability. The results indicated that high-transient heat flux up to 0.8 MW/m2 can cause sub-cooled boiling of the coolant in the first wall area of certain modules. Coolant returns to single phase soon after the end of the transient. According to the analysis, it is suggested that the blanket modules surrounding plasma have as similar structure design features as possible and sizes of the modules should be kept relatively small so as to obtain a reasonable pressure drop.     

中国氦冷固态实验包层模块In-box LOCA事故分析研究

中国氦冷固态实验包层模块(CN HCCB TBM)将在ITER 2号窗口进行测试,在测试期间,聚变中子和TBM内部材料发生核反应,产生氚和其他放射性物质。考虑到ITER的运行和工作人员与公众的安全,在进入ITER测试之前需要进行事故安全分析。本文应用MELOCR对HCCB TBM及其氦冷系统(HCS)进行建模,开展了TBM增殖区冷却板流道破口事故(In-box LOCA)安全研究,并对泄压罐体积,破口面积,隔离阀关闭延迟时间等关键参数进行敏感性分析。结果表明:在保守假设流道全破裂的工况下,box压力超过其压力限值4 MPa,而单根流道和5根流道破裂的工况下,box均未超过其压力限值;安装泄压罐和改变隔离阀关闭延迟时间能够有效的控制box压力。