聚变堆氦冷固态包层结构和6Li富集度对产氚率的影响

在聚变堆固态包层基本参数基础上，建立简化20°模型，包层分第1壁装甲、第1壁冷却板、氚增殖区和支撑结构。分别选择Li₄SiO₄和Li₂O做增殖材料，应用MCNP程序，研究第1壁结构布置和6Li富集度对产氚率的影响。结果表明：₆Li富集度适宜选择在30％~80％之间；第1壁选择Be装甲可提高产氚率；冷却管板的厚度应取3cm以下，以避免对产氚造成不利的影响。     

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。     

235U富集度的检测

采用γ能谱法和PC/FRAM软件分析检测铀样品中²³⁵U的富集度。实验结果表明,随机抽样的2个铀样品中,²³⁵U富集度测定值与标称值间相对偏差均在3%以内。该方法可用于铀样品中²³⁵U富集度的测量,并为今后核材料的测量提供一定的参考。     

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。     

氚增殖剂Li4SiO4 陶瓷小球的制备工艺

欧洲和中国聚变堆固态产氚包层（TBM）的氚增殖剂倾向于采用直径0.5～2mm的Li₄SiO₄陶瓷小球填充床。本工作探讨锂陶瓷小球的性能指标设计,研究挤压-滚圆、烧结法制备Li₄SiO₄小球的工艺可行性,测试分析小球的密度、直径、球形度、晶粒尺寸、压碎载荷等性能。研究表明：挤压-滚圆成型、1050℃无压烧结的Li₄SiO₄陶瓷小球密度为90.4%TD,堆积密度为52.9%TD;平均直径为0.95mm,标准偏差为0.15mm;球形度为1.10;平均压碎载荷为18.50N,标准偏差为2.76N;平均晶粒尺寸为14μm;相结构由Li₄SiO₄主晶相、少量Li₂SiO₃和Li₂Si₂O₅等组成。采用优化的挤压 滚圆、烧结工艺可制备出合格的Li₄SiO₄陶瓷小球产品。     

HFETR 6LiD中子转换器芯体厚度优化研究

介绍了通过在裂变堆中布置⁶LiD转换器将热中子转换为14MeV中子的原理,给出了⁶LiD转换器结构和材料的选取原则,建立了芯体厚度优化的目标函数。分别计算分析了在水和氦气两种冷却剂情况下,转换器结构、辐照样品数量对⁶LiD芯体厚度优化值的影响,给出⁶LiD芯体厚度对堆芯性能的影响。结果表明,布置于HFETR辐照孔道中的⁶LiD转换器芯体厚度优化设计值为0.7mm。     

加速器屏蔽体中氚的提取和比活度测量

使用红外炉发展了一种从活化混凝土中提取氚的方法。通过研究氚的提取效率与加热温度、加热时间的关系,获得了氚的提取优化条件为在800℃下加热30min,Ar气流量设置为200mL/min。氚由两个冷凝管收集,然后装到1个小玻璃瓶中,与液体闪烁体混合,使用液体闪烁计数器计数。使用这种方法提取氚只需55min。相对于使用电子炉提取氚的方法,氚的提取效率可达到100%。收集了两个加速器屏蔽体中的混凝土样品,测量了样品中的γ射线核素和氚的放射性比活度。结果显示,氚的比活度与¹⁵²Eu和⁶⁰Co的比活度有很强的相关性。可得出结论：氚主要产生于热中子俘获反应⁶Li(n,α)³H。因而,可使用⁶⁰Co的放射性比活度简单估计活化的混凝土样品中氚的放射性比活度。     

中国氦冷固态增殖剂实验包层系统氚输运初步分析

氚输运分析是开展中国氦冷固态增殖剂实验包层系统安全分析及未来聚变堆氚自持运行的重要研究内容之一。基于氚输运理论和固态增殖剂包层系统设计,利用FDS凤麟核能团队开发的聚变系统氚分析程序TAS,构建了固态增殖剂包层系统氚输运分析系统动力学模型。该模型氚输运结果与文献报道的吻合得很好,误差小于6%,验证了模型的正确性。针对中国氦冷固态增殖剂实验包层系统氚输运问题进行了两种计算方法(稳态、脉冲模式)的初步分析,获得了氚提取系统、氦气冷却系统回路氚分压,实验包层模块冷却流道、窗口室内氚提取系统和氦气冷却系统回路材料中氚滞留量,窗口室内氚提取系统和氦气冷却系统回路氚日渗透量等数据。最终对比结果显示,脉冲模式分析方法能够实时地跟踪源项的快速变化,更符合中国氦冷固态增殖剂实验包层系统实际运行情况。窗口室内氦气冷却系统回路材料中氚滞留量占到日产氚量的31.3%,因此需要在这些氚滞留损失严重的部位考虑适当的阻氚措施。     

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。     

6He(p,γ)7Li反应的实验设计与束流准备

在HI-13串列加速器次级束流线上,使用7Li初级束,通过²H(⁷Li,⁶He)³He反应产生了能量为26、32和38MeV的放射性6He次级束,纯化并准直后,6He次级束流纯度可达99%,强度最大为900s^-1。使用该次级束可实现对²H(⁶He,⁷Li)n反应的角分布测量,并结合理论计算,间接导出⁶He(p,γ)⁷Li的天体物理S因子和反应率。     

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.     

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.     

Status and progress of Indian LLCB test blanket systems for ITER

The lead–lithium ceramic breeder (LLCB) TBM and its auxiliary systems are being developed by India for testing in ITER machine. The LLCB TBM consists of lithium titanate as ceramic breeder (CB) material in the form of packed pebble beds. The FW structural material is ferritic martensitic steel cooled by high-pressure helium gas and lead–lithium eutectic (Pb–Li) flowing separately around the ceramic breeder pebble bed to extract the nuclear heat from the CB zones. Low-pressure helium is purged inside the CB zone for in situ extraction of bred tritium. Currently the LLCB blanket design optimization is under progress. The performance of tritium breeding and high-grade heat extraction is being evaluated by neutronic analysis and thermal–hydraulic calculations for different LLCB cooling configurations and geometrical design variants. The LLCB TBM auxiliary systems such as, helium cooling system (HCS), lead–lithium cooling system (LLCS), tritium extraction system (TES) process design are under progress. Safety analysis of the LLCB test blanket system (TBS) is under progress for the contribution to preliminary safety report of ITER-TBMs. This paper will present the status of the LLCB TBM design, process integration design (PID) of the auxiliary systems and preliminary safety analysis results.     

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.     

~2H(~6He,~7Li)n反应角分布和~7Li的质子谱因子

在中国原子能科学研究院HI-13串列加速器次级束流线上,使用6He次级束首次测量了质心系能量为9.1 MeV的2H(6He,7Li)n反应角分布,并用扭曲波波恩近似(DWBA)进行理论分析,导出了7Li的质子谱因子为0.40±0.02。     

Neutronic analysis of Indian helium-cooled solid breeder tritium breeding module for testing in ITER

India has proposed the helium-cooled solid breeder blanket concept as a tritium breeding module to be tested in ITER. The module has lithium titanate for tritium breeding and beryllium for neutron multiplication. Beryllium also enhances tritium breeding. A design for the module is prepared for detailed analysis. Neutronic analysis is performed to assess the tritium breeding rate, neutron distribution, and heat distribution in the module. The tritium production distribution in submodules is evaluated to support the tritium transport analysis. The tritium breeding density in the radial direction of the module is also assessed for further optimization of the design. The heat deposition profile of the entire module is generated to support the heat removal circuit design. The estimated neutron spectrum in the radial direction also provides a more in-depth picture of the nuclear interactions inside the material zones. The total tritium produced in the HCSB module is around 13.87 mg per full day of operation of ITER, considering the 400 s ON time and 1400 s dwell time. The estimated nuclear heat load on the entire module is around 474 kW, which will be removed by the high-pressure helium cooling circuit. The heat deposition in the test blanket model (TBM) is huge (around 9 GJ) for an entire day of operation of ITER, which demonstrates the scale of power that can be produced through a fusion reactor blanket. As per the Brayton cycle, it is equivalent to 3.6 GJ of electrical energy. In terms of power production, this would be around 1655 MWh annually. The evaluation is carried out using the MCNP5 Monte Carlo radiation transport code and FEDNL 2.1 nuclear cross section data. The HCSB TBM neutronic performance demonstrates the tritium production capability and high heat deposition.     

Effects of lithium burn-up on TBR in DEMO reactor SlimCS

Lithium in a breeding blanket is burned up through neutron nuclear reactions in fusion DEMO reactors. Effects of decrease of solid breeder materials due to lithium burn-up on tritium breeding ratio (TBR) are not systematically calculated in the past. For the SlimCS blanket design, TBR is calculated taking into account the lithium burn-ups by one dimensional Sn radiation transport calculation code ANISN in this study. The ⁶Li burn-ups are 8–79% after 10-year operation. TBR due to ⁶Li decreases to 40% of the initial one in some layer, while it increases in some layers. The TBR integrated over all the blanket decreases to around 96% of the initial one. The study makes it clear that the reduction of the TBR due to the lithium burn-up is not so large.