FEB—E氚循环系统的计算机模拟

对聚变实验增殖堆（ＦＥＢ）工程概要设计的氚燃料循环构造了一个动态子系统模型，研制了模拟氚燃料循环系统的计算机程序ＳＷＩＴＲＩＭ，计算运行一年后１０个子系统中的氚投料量和整个推系统总的氚投料量，这对预示起动一个聚变热功率的１５０ＭＷ量级的实验增殖堆所需的最少初始氚投料量有参考价值，计算结果表明，要求的最少初氚贮备量除了与燃料气体净化系统和同位互分离系统中氚的平均逗留时间有关外，还与包层液态锂中提取氚     

中国聚变工程实验堆启动氚投料量与氚平衡分析

根据中国聚变工程实验堆(CFETR)堆芯设计参数及燃料系统流程模型，采用平均停留时间方法，建立燃料循环子系统的氚转移模型用来描述氚在各子系统之间的输运、滞留等过程。采用该模型，分析了不同聚变功率水平、运行因子以及燃烧率对中国聚变工程实验堆的氚平衡以及启动氚投料量的影响。     

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研制出了用于计算氚投料量在FEB聚变堆各个子系统中的分布及其随时间变化的数值模拟程序包SWITRIM。通过近5年的使用，表明其运行良好、计算结果可靠。用SWITRIM数值模拟研究了聚变堆起动过程中的“氚坑深度和氚坑时间”新现象。简单介绍了SWITRIM程序包的组成和用户使用说明以及最新的运用等。     

聚变堆氚系统设计中的一些重要问题研究(Ⅱ)

在文献[1]中,计算了FEB-E 聚变堆PFC 材料内的氚滞留量、堆系统总的氚投料量、启动运行开始阶段的氚坑深度和氚坑时间大小。这里将讨论在ITER 的TBM 氚增殖包层内固体氚增殖剂中的氚如何高效率地被载氚气体带出并且以高效率地提取回收。本部分将进行创新的探索性研究并且提出某些减少氚滞留量和改善氚提取回收效率的新方案,例如：基于氘饱和的海绵效应;第一壁表面建立氘和铍的伴同沉积层;基于在低频外电场作用下载氚气分子和硅酸锂颗粒电极化旋转催化同位素交换速率的增强载氚气提取氚效率“SPB 方法”。     

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In part one published in the last issue, the tritium retention and the total tritium inventory in PFC materials of FEB-E fusion reactor had been calculated. The tritium well depth, tritium well time during the FEB-E fusion reactor start-up and initial operation phase had been obtained. In this part, how to improve tritium recovery efficiency in the ITER TBM solid breeder blanket with using purge gas has been discussed. Some new innovative schemes for reducing tritium retention and improving tritium recovery efficiency are proposed. Such as, sponge mechanism based on deuterium saturated PFC materials; deuterium and beryllium co-deposition layer created on first wall surface; SPB scheme for enhancing tritium recovery efficiency of purge gas in ceramic breeder blanket based on the electrical polarization rotations catalyzing isotope exchange rate enhancement resulted from applied low frequency electric-field, of Li4SiO4 grain and purge gas molecular particles and so on, are explored.     

锂铅合金释氚实验研究

由于锂铅合金因具有高增殖比、低活泼性和可能作为冷却剂的特点,被认为是最有潜力的能源堆包层氚增殖材料。在理论模型描述熔融锂铅合金氚释放行为的基础上,开展了中子辐照后Li₁₇Pb₈₃合金的离线氚释放实验。结果表明: 释放氚的化学形式99％以上为难溶于水的成分（HT或T2）; 氚滞留时间随载气中氢分压的增加而减小,氢分压达到1000 Pa后变为常数,且与实体积无关;氚释放速率对温度的依赖性符合Arrhenius定律。以此为基础得到的氚在熔融锂铅中的动力学参数结果,虽与文献值有差异,但同样证明了在633—973 K的范围内, 氚从液态锂铅到气相的整个释放过程中起决定作用的是氚在合金内的扩散和气\|液界面的多相反应重组。Lithium\|lead alloy is considered to be one of the most prominent tritium breeding materials for the fusion reactor blanket because of its high breeding ratio, and low reactivity and possible use as coolant. An out\|of\|pile experiment of tritium release from Li₁₇Pb₈₃ alloy was performed after neutron irradiation on the base of mathematical model to describe tritium release behavior from an eutectic lithium\|lead alloy. The results suggest that the dominant chemical form of the released tritium (>99%) was the water\| insoluble component (HT or T2). Tritium residence time decreased with increasing H₂pressure in carrier gas up to 1000 Pa, and above this concentration limit it became constant and not influenced by the plenum volume. The temperature dependence of the tritium release rate can be described by an Arrhenius law. Consequently, the present results on the kinetic parameters of tritium in molten Li₁₇Pb₈₃alloy are considered to be different from the values in literature, but it is the same that the overall release process is governed by the diffusion of tritium atoms in the Li₁₇Pb₈₃and by the heterogeneous reaction at the gas\|eutectic interface of the tritium atom recombination at temperatures from 633 to 973 K.     

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聚变堆第一壁表面和PFC材料内的氚滞留量、堆系统总的氚投料量多高？在启动和运行的开始阶段的氚坑深度,氚坑时间的大小是多少？在TBM氚增殖包层内固体氚增殖剂中的氚能否高效率地被载氚气体带出来并且以高效率地提取回收？能否找到某些新机制解决这些问题是决定实现ITER的预期目标和最终实现聚变能的实际运用成败的关键问题。本文第(Ⅰ)部分回答前面两个问题,在下期第(Ⅱ)部分将进行创新的探索性研究并且提出某些减少氚滞留量和改善氚提取回收效率的新方案,例如：基于氘饱和的海绵效应;第一壁表面建立氘和铍的伴同沉积层;基于在低频外电场作用下载氚气分子和硅酸锂颗粒电极化旋转催化同位素交换速率增强提高载氚气提取氚效率“SPB方法”等等。     

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Based on the design of the 2015 version of China Fusion Engineering Test Reactor (CFETR) water cooled ceramic breeder (WCCB) blanket modules surrounding the plasma, a tritium transport model has been developed. Tritium transport analysis has been carried out for each blanket module with different breeding zones, purge gas loop, coolant loop and steam generator. The results indicate that the concentration, permeability and retention of tritium among blanket modules are different. For all of the WCCB blanket modules in CFETR, the tritium retention inside the breeder is 6.62×10^-2g, the tritium retention inside the structural materials is 2.01g, the tritium retention inside purge gas and coolant loop are 4.03×10^-4g and 0.19g respectively, the tritium permeation through the steam generator tube walls is 20mg•y^-1, the tritium permeation from the coolant pipes is 0.1mg•y^-1.     

增殖剂球床载气体流动特性

增殖剂球床是聚变堆或混合堆产氚包层可选结构之一,准确把握增殖剂球床中载带气体的流动特性有助于提高对球床载氚过程的认识并优化包层设计。采用离散元程序PFC3D模拟增殖剂小球的填充行为,在球床内不同位置随机截取不同尺寸的控制体,利用布尔运算中的"差集"得到孔隙范围,建立孔隙分布的三维几何模型,进一步划分网格并用计算流体力学(CFD)方法求解,得出控制体上单位长度的压降以及单元体内的速度分布特征,计算结果发现载带气体速度分布与γ分布很类似,且只要选取恰当的控制体,通过计算流体力学方法可以较好地分析整个球床孔隙内流体的流动,有利于进一步研究载氚及相关过程。     

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水冷陶瓷包层是中国聚变工程实验堆(CFETR)的三种候选包层概念之一。基于CFETR水冷陶瓷包层的一维中子学模型,通过蒙特卡罗输运模拟程序MCNP和活化计算程序FISPACT的耦合计算,经三维转换系数修正,分析了CFETR水冷陶瓷包层时间相关产氚特性。结果表明,当CFETR运行因子为0.5,聚变功率为200MW时,水冷陶瓷包层在运行5年、10年、20年后,氚增殖率(TBR)的降低都不显著,但是年产氚剩余量的降低很明显。此外,产氚包层内初始时刻TBR对产氚特性的影响也很大。     

CFETR 水冷陶瓷包层时间相关产氚特性初步分析

水冷陶瓷包层是中国聚变工程实验堆(CFETR)的三种候选包层概念之一。基于CFETR水冷陶瓷包层的一维中子学模型,通过蒙特卡罗输运模拟程序MCNP和活化计算程序FISPACT的耦合计算,经三维转换系数修正,分析了CFETR水冷陶瓷包层时间相关产氚特性。结果表明,当CFETR运行因子为0.5,聚变功率为200MW时,水冷陶瓷包层在运行5年、10年、20年后,氚增殖率(TBR)的降低都不显著,但是年产氚剩余量的降低很明显。此外,产氚包层内初始时刻TBR对产氚特性的影响也很大。     

聚变示范堆气体注入系统氚渗透分析

以中国聚变工程实验堆(CFETR)的堆芯参数为例, 参照ITER 的相关工程设计,对聚变示范堆(DEMO) 气体注入系统(GIS)作出了初步的管道布局考虑。基于此布局,计算了稳态运行下GIS 含氚管道气流压降,对燃料储存与输运系统的泵压头提出了要求,得出相应泵压头下GIS 的氚渗透量。计算结果表明,重力势是管道气流压降的主要原因,在负压情况下,GIS 的氚渗透对整体氚循环的影响不大。这为研究DEMO 的氚自持问题提供了相关参数,为GIS歧管布局的进一步优化奠定了基础。     

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In the present work, tritium permeation issue of gas injection system (GIS) in DEMO were focused. Based on the reactor core parameters of Chian Fusion Engineering Test Reactor (CFETR) and referring to accumulated experience of GIS design in ITER, preliminary pipes layout of DEMO GIS was performed. Steady-state tritium gas pressure drop calculation along the pipes and pump head request to the fuel storage and delivery system were made on the basis of the GIS’s structure layout. Tritium permeation under corresponding pump head was calculated at the same time. According to the calculation results, gravitational potential is the main determinant to pressure drop in pipe flow and the tritium permeation in GIS is of little amount to the whole tritium cycle in the case negative pressure. So this work offers related parameters to tritium self-sufficiency issues in DEMO, and also lays the foundation for further optimization of GIS’s structure layout.     

球形托卡马克聚变嬗变堆中子学设计

基于对球形托卡马克ST聚变堆的研究,提出了ST聚变嬗变堆的设计概念。对堆芯参数作了初步选择,确定了一组适合于嬗变包层的堆芯参数供中子学计算和结构设计参考,给出了在以嬗变次锕系元素（MA）核废物为目标的一维中子学计算结果。     

Fuel cycle for a fusion neutron source

The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion–fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium–tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m³Pa/s, and temperature of reactor elements up to 650°C). The deuterium–tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.     

随机填充增殖剂球床内载气流动特性数值模拟

固态氚增殖包层是聚变堆及聚变-裂变混合堆产氚包层的重要候选结构之一,其球床通道内载气流动特性将影响氚提取效率。利用离散元方法(DEM)生成随机填充增殖剂球床,通过径向孔隙率分布验证其合理性,计算流体力学(CFD)模拟计算其通道内气体流场特征。模拟得到:球床内吹扫氦气流速随孔隙率波动并随入口流速增大而均匀增大,通道内氦气流向及流速变化显著,Blake-Kozeny方程可良好预测该随机填充球床通道压降。     

Z箍缩驱动聚变-裂变混合堆氚燃料循环系统氚盘存量与氚自持初步分析

为研究氚自持条件,建立了Z-FFR氚分析模型,基于理论方程和氚平均滞留时间方法进行计算,得到稳态运行时排灰气处理系统、氚增殖提取系统、同位素分离系统、水去氚化系统的氚质量流分别为52.30,25.40,81.30,3.60 g/day,对应的氚盘存量为52.30,25.40,8.13,1.80 g。同时以氚质量流推导出氚自持判断条件,分析了设计参数能够满足氚自持要求,同时获得了燃烧效率、氚增殖率、提取效率与氚自持的互补关系,三者作为关键参数相互依存,于临界值、设计值、理想值之间分析了氚的自持情况。     

紧凑型托卡马克聚变实验堆包层设计及中子学分析

利用蒙特卡罗粒子输运程序MCNP,对紧凑型托卡马克(CT)聚变实验堆方案进行了三维中子学计算分析,特别是氚增殖比(TBR)计算。包层采用了氦冷固态包层(HCCB Blanket)概念,选用正硅酸锂(Li₄SiO₄)作为氚增殖剂,铍作为中子倍增剂,低活化铁素体钢为结构材料。给出了不同增殖单元在不同的排列方式下,TBR 随增殖单元厚度的变化情况,得到了满足氚自持条件下的初步优化的包层设计方案,给出了相关中子壁负载、能量沉积和功率密度的相关特性结果,为后续热工水力和相关设计分析提供了数据支持。     

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Using Monte Carlo particle transport code MCNP, the 3D neutronics analysis as well as TBR calculation for high power compact tokamak (CT) test reactor were done. The structure design for helium cooling ceramic breeder (HCCB) blanket was carried out with Li₄SiO₄, Be and the low activated ferrite steel as the tritium breeder, neutron multiplier and structural material respectively. The variation of TBR vs the breeding unit thickness was simulated, under different breeding unit with different structure arrangement. The preliminary optimum design of blanket with tritium self-sufficiency was obtained. The related characteristics of neutron wall loading, energy deposit and power density was given, as supplies data support for the thermal hydraulics and related design and analysis in next step.     

正硅酸锂陶瓷小球离线释氚实验研究

中国氦冷球床包层(CH HCSB TBM)初步采用Li₄SiO₄陶瓷小球作为氚增殖材料,实验研究了Li₄SiO₄陶瓷小球的中子辐照产氚性能。将冷冻成型法制备的Li₄SiO₄ 陶瓷小球置于反应堆中辐照100min,然后在离线释氚实验平台上进行退火行为研究。实验结果表明,在用He +1% H₂为载气,流速为100mL•min^-1,升温速率为5K•min^-1的实验条件下,氚气(HT+T₂)是Li₄SiO₄陶瓷小球的主要释氚形态,占总氚的70%左右(不包括自由氚中的氚气),在400℃~700℃范围内出现两次释氚峰;氚化水(HTO+T₂O)所占比例小于20%,主要在300℃~500℃的低温段进行释放;氚在800℃前基本释放完,小球退火后的残氚量小于1%。冷冻成型干燥法制备的Li₄SiO₄陶瓷小球在300℃~700℃范围内有较好的释氚性能,氚残留量低,在聚变堆固态氚增殖包层设计中具有一定应用价值。