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。     

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。     

高氚增殖比包层的设计及热工水力分析

本文设计了一种高氚增殖比包层(HBRB),该包层采用多孔U-10Zr合金作为中子倍增剂,Li4SiO4球床作为增殖剂,低活化马氏体(RAFM)钢作为结构材料。在详细研究包层加工工艺、流量分配、中子性能等问题的基础上,完成了包层内部详细结构设计。利用中子学软件分析计算了包层的氚增殖比(TBR)和热沉积分布,并根据计算结果对包层进行热力耦合分析。结果表明:包层TBR较高,且核性能稳定;冷却剂的流量分配情况和压降合理;包层内各组件冷却充分,温度和结构材料热应力不超过限值。     

聚变裂变混合发电堆水冷包层中子学设计分析

主要针对聚变裂变混合发电堆FDS-EM水冷包层的能量倍增因子M和氚增殖率TBR等中子学参数进行优化计算。FDS-EM包层主要设计目标是在氚自持的基础上获得约1 GW的电功率,并且尽可能长时间连续运行不换料。通过初步设计分析给出一个使用核废料(压水堆卸出的废料钚、锕系加上贫铀)作为裂变燃料,能够实现氚自持、能量倍增因子约为90等设计目标,且连续运行至少10年不换料的中子学方案。     

中国氦冷固态增殖剂实验包层模块材料研究进展

在未来核聚变反应堆中,为补充氚的消耗,需要在核聚变堆的包层中进行氚的在线增殖,以维持核聚变反应的持续进行。为验证这一关键技术,在国际热核聚变实验堆(ITER)上开展了ITER TBM计划(实验包层项目)。作为ITER计划成员方之一,中方以中国氦冷固态增殖剂实验包层模块(HCCB TBM)概念参与ITER TBM计划。HCCB TBM现今进入初步设计阶段,而材料的制备技术和性能数据是支撑其结构设计、安全分析和服役工况评估的基础。本文综述和分析了HCCB TBM结构材料低活化铁素体/马氏体钢(RAFM钢)与功能材料氚增殖剂和中子倍增剂的研究现状,并对这些材料下一步的研究方向进行了展望。     

中国氦冷固态增殖剂实验包层模块材料研究进展

在未来核聚变反应堆中,为补充氚的消耗,需要在核聚变堆的包层中进行氚的在线增殖,以维持核聚变反应的持续进行。为验证这一关键技术,在国际热核聚变实验堆（ITER）上开展了ITER TBM计划（实验包层项目）。作为ITER计划成员方之一,中方以中国氦冷固态增殖剂实验包层模块（HCCB TBM）概念参与ITER TBM计划。HCCB TBM现今进入初步设计阶段,而材料的制备技术和性能数据是支撑其结构设计、安全分析和服役工况评估的基础。本文综述和分析了HCCB TBM结构材料低活化铁素体/马氏体钢（RAFM钢）与功能材料氚增殖剂和中子倍增剂的研究现状,并对这些材料下一步的研究方向进行了展望。     

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。     

中国氦冷固态实验包层2×6模块中子学计算分析

为了满足ITER对波纹度的要求,核工业西南物理研究院提出了新的减少低活化铁素体钢的氦冷固态（HCSB）实验包层模块（TBM）设计方案。采用MCNP程序及ITER全堆MCNP模型,对新设计的2×6HCSB-TBM进行三维中子学计算分析,给出了模块产氚率、核热沉积和功率密度分布等结果。在ITER运行因子为22%时,HCSB-TBM的产氚率为12.68mg/d。TBM内总核热沉积为522.5kW,最高功率密度为11.8W/cm³,出现在氚增殖区Li₄SiO₄中。计算结果可为TBM进一步的结构、热工水力学优化及其他系统设计提供中子学数据。     

氦气、水和熔盐冷却的混合堆中子学性能研究

氦气、水、熔盐(Flibe)在强磁场中流动不存在严重的MHD问题,因此适合在基于磁约束的聚变-裂变混合堆中作为冷却剂.针对氦气、水、Flibe这3种冷却剂对混合堆包层中子学性能的影响进行研究,分析包层中能谱特点及燃料增殖特性.通过燃耗计算,研究氚增殖率(TBR)、能量倍增因子(M)、keff等随运行时间的变化.中子学输运采用三维蒙特卡罗程序MCNP.计算结果表明,不同的冷却剂对混合堆系统中子能谱影响很大:氦冷系统的能谱最硬,主要发生快中子裂变,氚增殖效果最好;水冷系统的能谱最软,产能最多,但需提高TBR;Flibe冷系统的能谱较硬,产能最少.     

氚增殖剂球床组件堆内辐照物理热工计算分析

为验证在中国先进研究堆(CARR)内进行国际热核聚变实验堆(ITER)氚增殖包层模块(TBM)辐照实验的可行性和安全性,进行了氚增殖剂球床组件堆内辐照物理及热工计算分析。氚增殖剂包层模块主要是固态氚增殖剂陶瓷球床。本文采用Monte Carlo粒子输运模拟程序对氚增殖剂球床进行堆内建模,计算球床的中子注量率、能量沉积和产额,得到不同功率下球床的中子注量率、发热功率和产氚速率以及球床组件引入反应堆的反应性。根据物理计算得到的组件各部件发热情况建立热工计算一维模型,通过更改反应堆功率得到满足实验要求的工况并采用三维程序进行验证。物理与热工计算分析的结果表明,在反应堆运行功率为20 MW的工况下球床组件各部件的温度均不超过限值。     

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.     

Conceptual design of a water cooled breeder blanket for CFETR

China Fusion Engineering Test Reactor (CFETR) is an ITER-like superconducting tokamak reactor. Its major radius is 5.7 m, minor radius is 1.6 m and elongation ratio is 1.8. Its mission is to achieve 50–200 MW of fusion power, 30–50% of duty time factor, and tritium breeding ratio not less than 1.2 to ensure the self-sufficiency. As one of the breeding blanket candidates for CFETR, a water cooled breeder blanket with superheated steam is proposed and its conceptual design is being carried out. In this design, sub-cooling water at 265 °C under the pressure of 7 MPa is fed into cooling plates in breeding zone and is heated up to 285 °C with saturated steam generated, and then this steam is pre-superheated up to 310 °C in first wall (FW), final, the pre-superheated steam coming from several blankets is fed into the other one blanket to superheat again up to 517 °C. Due to low density of superheated steam, it has negligible impact on neutron absorption by coolant in FW so that the high energy neutrons entering into breeder zone moderated by water in cooling plate help enhance tritium breeding by ⁶Li(n,α)T reaction. Li₂TiO₃ pebbles and Be₁₂Ti pebbles are chosen as tritium breeder and neutron multiplier respectively, because Li₂TiO₃ and Be₁₂Ti are expected to have better chemical stability and compatibility with water in high temperature. However, Be₁₂Ti may lead to a reduction in tritium breeding ratio (TBR). Furthermore, a spot of sintered Be plate is used to improve neutron multiplying capacity in a multi-layer structure. As one alternative option, in spite of lower TBR, Pb is taken into account to replace Be plate in viewpoint of safety. In this contribution, study on neutronics and thermal design for a water cooled breeder blanket with superheated steam is reported.     

聚变堆交叉冷却固态包层中子学设计优化

针对聚变堆固态包层设计路线,提出了一个交叉排列氦冷固态包层概念。设计采用Be、Li₂TiO₃分层球床。两种尺寸的氦气冷却管道交叉排列,分两个回路同时冷却,以增加系统安全可靠性。分析比较了4种⁶Li富集度布置方案。结果表明：径向远离第一壁降低⁶Li富集度较为合理,靠近第一壁的增殖层⁶Li富集度不能过低,以减少长期运行中Li的消耗对氚增殖性能的影响。借助蒙特卡罗程序MCNP建立11.25°对称模型,全堆包层氚增殖率为1.176,包层寿期内产氚性能稳定,在包层寿命运行时间内的燃耗分布相对均匀。     

Progress on design and R&D for helium-cooled ceramic breeder TBM in China

The helium-cooled ceramic breeder (HCCB) test blanket module (TBM) is the primary option of the Chinese TBM program. Current progress on the design and R&D for Chinese helium-cooled ceramic breeder TBM (CN HCCB TBM) in China is presented. The main updated design and related R&D of CN HCCB TBM are introduced briefly. The mock-up fabrication and component tests for Chinese test blanket module are being carried out. Recent status of the components and fabrication technology development is also reported. The neutron multiplier Be pebbles, tritium breeder Li₄SiO₄ pebbles, and structure material CFL-1 are being prepared in the laboratory scale. The fabrication of 1/3 sized mock-up and construction of a He test loop are being carried out. The key technology development is proceeding to the large scale mock-up fabrication and demonstration tests toward on ITER testing.     

聚变堆混和球床包层中子学和热工水力特性研究

在聚变堆初步概念设计的基础上，针对固态包层设计路线，提出了一个先进的氦冷固态包层概念。设计采用Be₁₂Ti和Li₂TiO₃陶瓷小球混和球床，物理和化学相容性好；采用SiC作为结构材料，提高耐高温性能及氦气出口温度。计算结果表明：选择Be₁₂Ti和Li^₂TiO₃球体积比在2和4之间较合理；在Be₁₂Ti和Li₂TiO₃球体积比为3时，⁶Li富集度取30％~80％较适宜；球床的最高温度低于材料的温度限值，温度分布合理均匀。该方案可较大程度提高热效率和改善中子学以及氚增殖性能。     

Preliminary performance analysis and optimization based on 1D neutronics model for Indian DEMO HCCB blanket

India, under its breeding blanket R&D program for DEMO, is focusing on the development of two tritium breeding blanket concepts; namely the lead-lithium-cooled ceramic breeder and the helium-cooled ceramic breeder (HCCB). The study presented in this paper focuses on the neutronic design analysis and optimization from the tritium breeding perspective of the HCCB blanket. The Indian concept has an edge-on configuration and is one of the variants of the helium-cooled solid breeder blanket concepts proposed by several partner countries in ITER. The Indian HCCB blanket having lithium titanate (Li₂TiO₃) as the tritium breeder and beryllium (Be) as the neutron multiplier with reduced-activation ferritic/martensitic steel structure aims at utilizing the low-energy neutrons at the rear part of the blanket. The aim of the optimization study is to minimize the radial blanket thickness while ensuring tritium self-sufficiency and provide data for further neutronic design and thermal-hydraulic layout of the HCCB blanket. It is found that inboard and outboard blanket thicknesses of 40 cm and 60 cm, respectively, can give a tritium breeding ratio (TBR) >1.3, with 60% ⁶Li enrichment, which is assumed to be sufficient to cover potential tritium losses and associated uncertainties. The results also demonstrated that the Be packing fraction (PF) has a more profound impact on the TBR as compared to 6Li enrichment and the PF of Li₂TiO₃.     

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.