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。     

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.     

Updated neutronics analyses of a water cooled ceramic breeder blanket for the CFETR

The water cooled ceramic breeder (WCCB) blanket employing pressurized water as a coolant is one of the breeding blanket candidates for the China Fusion Engineering Test Reactor (CFETR).Some updating of neutronics analyses was needed,because there were changes in the neutronics performance of the blanket as several significant modifications and improvements have been adopted for the WCCB blanket,including the optimization of radial build-up and customized structure for each blanket module.A 22.5 degree toroidal symmetrical torus sector 3D neutronics model containing the updated design of the WCCB blanket modules was developed for the neutronics analyses.The tritium breeding capability,nuclear heating power,radiation damage,and decay heat were calculated by the MCNP and FISPACT code.The results show that the packing factor and 6Li enrichment of the breeder should both be no less than 0.8 to ensure tritium self-sufficiency.The nuclear heating power of the blanket under 200 MW fusion power reaches 201.23 MW.The displacement per atom per full power year (FPY) of the plasma-facing component and first wall reach 0.90 and 2.60,respectively.The peak H production rate reaches 150.79 appm/FPY and the peak He production reaches 29.09 appm/FPY in blanket module # 3.The total decay heat of the blanket modules is 2.64 MW at 1 s after shutdown and the average decay heat density can reach 11.09 kW m-3 at that time.The decay heat density of the blanket modules slowly decreases to lower than 10 W m-3 in more than ten years.     

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.     

Impact Analysis of Helium Cooled Solid Blanket Structures on the Tritium Breeding Performance in CFETR

Chinese Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor to bridge the gap between ITER and future fusion power plant. As its objectives are to demonstrate generation of fusion power and to realize tritium self-sufficiency, the tritium breeding ratio (TBR) is a key design parameter. In the blanket design and optimization, the structures such as the first wall (FW), cooling plate (CP), stiffening plate (SP), cap and some other design parameters in detailed 3-D model have significant impacts on the tritium breeding performance. Based on a helium cooled solid breeder blanket option for CFETR, the impact analysis of the helium cooled solid blanket structures on tritium breeding performance was performed in this paper. Firstly, the detailed 3D neutronics model was built by using of a CAD to Monte Carlo Geometry conversion tool McCad. Then based on the detailed 3D neutronics model, the impact analyses of the blanket structures on tritium breeding performance were carried out, which include the FW, CP, SP, cap and side wall. By the sensitivity study of the blanket structures on the TBR, it gave the TBR variation trend and references for the blanket design and optimization.     

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。     

中国聚变工程试验堆包层的核热耦合效应研究

本文以中国聚变工程试验堆（CFETR）的氦冷固态包层和水冷固态包层为研究对象,基于蒙特卡罗程序MCNP和计算流体力学程序FLUENT,利用3D-1D-2D耦合方法和伪材料方法,分别对200 MW的氦冷固态包层和水冷固态包层及1.5 GW的水冷固态包层方案进行了核热耦合计算分析。研究结果表明,金属铍的热散射效应和轻水密度是聚变包层核热耦合效应的主要来源,核热耦合效应对氦冷固态包层的影响可忽略,对水冷固态包层的氚增殖比和温度分布有一定程度的影响。     

Activation and Environmental Aspects of In-Vacuum Vessel Components of CFETR

The water-cooled ceramic breeder(WCCB) blanket is one of the three candidates of China's Fusion Engineering Test Reactor(CFETR). The evaluation of the radioactivity and decay heat produced by neutrons for the in-vacuum vessel components is essential for the assessment of radioactive wastes and the safety of CFETR. The activation calculation of CFETR in-vacuum vessel components was carried out by using the Monte Carlo N-Particle Transport Code MCNP, IAEA Fusion Evaluated Nuclear Data Library FENDL2.1, and the nuclear inventory code FISPACT-2007 and corresponding EAF-2007 libraries. In these analyses, the three-dimensional(3-D) neutronics model was employed and the WCCB blanket, the divertor, and the shield were modeled in detail to provide the detailed spatial distribution of the neutron flux and energy spectra. Then the neutron flux, energy spectra and the materials specification were transferred to FISPACT for the activation calculation with an assumed irradiation scenario of CFETR. This paper presents the main results of the activation analysis to evaluate the radioactivity, the decay heat, the contact dose, and the waste classification of the radioactive materials. At the time of shutdown, the activity of the WCCB blanket is 1.88×10~(19)Bq and the specific activity, the decay heat and the contact dose rate are 1.7×10~(13)Bq/kg, 3.05 MW, and 2.0×10~3Sv/h respectively. After cooling for 100 years, 79%(4166.4 tons) radioactive wastes produced from the blanket, divertor,high temperature shield(HTS) and low temperature shield(LTS) need near surface disposal, while21%(1112.3 tons) need geological disposal. According to results of the contact dose rate, all the components of the blanket, divertor, HTS and LTS could potentially be recycled after shutdown by using advanced remote handling equipment. In addition, the selection of Eurofer97 or RAFM for the divertor is better than that of SS316 because SS316 makes the activity of the divertor-body keep at a relatively high level.     

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。     

Investigation of the packing structure of pebble beds by DEM for CFETR WCCB

Li₂TiO₃/Be₁₂Ti mixed pebble beds with multi-sized particles are one of the potential candidates for the WCCB (water-cooled ceramic breeder blanket) of the CFETR (China Fusion Engineering Test Reactor). To meet the neutronics requirements of a WCCB, a study of the packing structure of the concerned pebble bed is necessary. In this paper, the discrete element method (DEM) is applied to produce a prototypical blanket pebble bed by directly simulating the contact state of each individual particle using basic interaction laws. According to the current simulation, the packing factor of a mono-sized pebble bed is 0.62–0.64, while the value will become more than 0.75 for Li₂TiO₃/Be₁₂Ti mixed breeding pebble bed with a diameter ratio of not less than 5 as well as an appropriate mixed volume ratio, and thus can meet the neutronics requirements.     

cosRMC复杂曲面开发及在聚变中子学的应用

中国聚变工程实验堆（CFETR）是我国自主设计和研制的重大科学工程,CFETR旨在与ITER相衔接和补充,为研制DEMO级别聚变堆电站提供必要的技术。蒙特卡罗方法在聚变中子学与屏蔽设计等方面具有重要作用。本文基于自主化蒙特卡罗程序cosRMC,研究了蒙特卡罗复杂曲面建模的数学模型和计算方法,开发了复杂曲面建模功能,并通过PPCS（power plant conceptual study）模型验证了该功能实现的正确性。然后构建了CFETR的三维精细化模型,并利用该模型对CFETR包层设计中的关键中子学参数进行计算分析。结果表明,cosRMC对中子学参数氚增殖比、中子壁载荷和核热沉积的计算结果与MCNP的计算值吻合良好,相对偏差均小于5%,满足工程设计需求。研究证明了cosRMC应用于聚变堆包层中子学分析的正确性和有效性。CFETR中子学参数的计算分析,也为其设计和优化提供了参考。     

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T riSim-Dynamic,在此基础上进行修改完善,利用该程序对CFETR...     

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.     

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。     

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.     

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。     

超临界水冷固态实验包层中子学与热工水力特性研究

基于国际热核聚变实验堆(ITER)实验包层方案,提出了一个超临界水冷固态实验包层概念设计方案。设计采用Be作为中子倍增剂,Li4SiO4作为氚增殖剂,CLAM钢作为结构材料。包层第一壁采用多层盘道设计以提高第一壁出口温度,内部采用增殖剂与中子倍增剂分层布置以提高热沉积与氚增殖率。为验证包层设计的可行性,分析计算了三维包层氚增殖率与热沉积的分布,然后根据中子学计算得到的结果对超临界水冷固态实验包层进行了数值模拟研究。结果表明:包层功率密度分布较合理;氚增殖率满足运行中氚自持的要求;在冷却剂出口温度达到500℃条件下材料温度不超过限值。该设计方案能满足中子学设计与热工水力的要求。     

托卡马克聚变中子源的数值模型研究

托卡马克(Tokamak)聚变装置中子学分析中,聚变中子源描述是重要的输入参数,其准确性直接影响分析结果的可靠性。通过分析ITER和欧洲聚变示范堆(EU DEMO)中子学分析中所采用的聚变中子源模型,提出了一种完整描述Tokamak中L-mode、H-mode等离子体的D-D、D-T聚变中子源的数值模型。在中国聚变工程实验堆(CFETR)的工程集成设计平台上,编写了基于蒙特卡罗算法的程序SCG求解该数值模型,实现了读取(零维)等离子体参数、输出可供典型中子学软件MCNP直接读取的中子源定义文件的功能。以CFETR氦冷球床包层的中子学分析模型为基准,在相同的L-mode等离子体D-T聚变工况下,相较于采用EU DEMO源子程序,采用本模型计算得到的中子壁负载差异最大值为2.02%,包层氚增殖率差异为0.18%,全堆能量增益因子的差异为0.23%。结果表明,本模型与其他源描述的差异较小,可应用于CFETR的中子学分析。