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.     

Z箍缩聚变裂变混合堆包层中子学分析

作为一种有竞争力的能源系统,Z箍缩聚变裂变混合堆(Z-FFR)正在开展概念研究,包层研究正是其中重要的一部分。建立了Z-FFR包层设计模型,分析了包层影响因素、中子平衡、通量与功率密度、燃耗等方面,表明该包层设计在50年内能量放大因子、氚增殖比和燃料增殖比的平均值分别为14.91,1.294和5.140,满足设计要求。针对聚变源的脉冲特性进行了包层的瞬态中子学分析,发现燃料区中子脉冲可分为聚变中子、瞬发裂变中子和缓发裂变中子脉冲三个部分,绝大部分热量约在0.01s内沉积。结果较完整地给出了Z-FFR包层的中子学参数,为概念研究提供了基础。     

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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。     

Radiation effect of neutrons produced by D–D side reactions on a D–3He fusion reactor

One of the most important characteristics in D–³He fusion reactors is neutron production via D–D side reactions. The neutrons can activate structural material, degrading them and ultimately converting them into high-level radioactive waste, while it is really costly and difficult to remove them. The neutrons from a fusion reactor could also be used to make weapons-grade nuclear material, rendering such types of fusion reactors a serious proliferation hazard. A related problem is the presence of radioactive elements such as tritium in D–³He plasma, either as fuel for or as products of the nuclear reactions; substantial quantities of radioactive elements would not only pose a general health risk, but tritium in particular would also be another proliferation hazard. The problems of neutron radiation and radioactive element production are especially interconnected because both would result from the D–D side reaction. Therefore, the presentation approach for reducing neutrons via D–D nuclear side reactions in a D–³He fusion reactor is very important. For doing this research, energy losses and neutron power fraction in D–³He fusion reactors are investigated. Calculations show neutrons produced by the D–D nuclear side reaction could be reduced by changing to a more ³He-rich fuel mixture, but then the bremsstrahlung power loss fraction would increase in the D–³He fusion reactor.     

Fuel breeding in a hybrid muon catalyzed fusion reactor

We discuss the processes of nuclear fuel burnup and plutonium breeding in the uranium blanket of a hybrid mesocatalytic reactor. The time dependence of the nuclear fuel isotopic concentrations was calculated by the BURNFL code. Using a three-dimensional Monte Carlo MORSE-H code the plutonium and tritium breeding coefficients, the fission rates of uranium and plutonium and a specific power distribution in the blanket were computed. The total fusion energy multiplication factor was obtained as a function of the fuel residence time using results of a detailed calculation of the mesocatalytic channel and estimations of the electronuclear channel.     

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     

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。     

ITER驱动次临界包层总体结构概念设计

详细介绍了ITER驱动的次临界包层结构设计,沿环向360°整个次临界包层被分成36瓣,单瓣包层以等离子工作腔为分界面被分为内、外两部分,分别由第一壁结构、支承结构、燃料区结构、产氚区结构和锆包壳结构等组成。有别于ITER装置现有的小模块包层结构,单瓣内、外包层被设计成一种整体式内置结构,从而减少了裂变燃料区中大量内嵌冷却剂压力管道接头数量、缩短了换料周期并节约了成本。同时考虑到ITER装置本体结构空间对次临界包层的限制,提出了一种既能满足包层热工-流体要求又能实现包层工程焊接安装的管道汇总结构。最后运用Pro/e建模软件建立了包层三维CAD结构图,为后续结构力学分析输出了有限元计算模型。     

磁约束聚变堆次临界包层概念设计研究进展

为使磁约束聚变堆实现能量放大与氚自持,在其等离子体区周围设置次临界包层和产氚包层。采用天然铀合金燃料、轻水作冷却剂兼慢化剂,内嵌压力管式的次临界包层设计方案,通过对包层物理性能、结构概念设计、热工水力性能和安全分析,表明该方案可将聚变能量放大10倍以上,氚增殖比大于1.15,具有天然的临界安全性和良好余热安全性能。立足于近中期可利用的聚变技术,力争实现聚变能源的提前商用,为我国能源可持续发展提供一种有竞争力的技术选项。     

美国激光惯性约束聚变能源研究综述

简要地介绍了美国激光惯性约束聚变能源( LIFE ) 的研究现状与发展前景。基于美国国家点火装置( NIF ) 的近期进展,美国利弗莫尔实验室提出了激光惯性约束聚变能源设想,并开始了分解研究。设想用新型二极管泵浦固体激光器产生1.4~2.0 MJ 的激光能量,靶丸聚变增益25~30,打靶频率10~15Hz,实现350~500 MW聚变功率,相当于聚变中子源强1.3×1020 ~1.8×1020 n/s。以此驱动次临界裂变包层,使能量再倍增4~10 倍,实现1 GW电功率的输出。采用创新设计的燃料元件,包层可达到90%以上的燃耗深度,形成一个安全、无碳、燃料资源丰富、核废料少、可持续发展的新型核能源系统。In this paper the present study situation and prospect of the American laser-based Inertial Confinement Fusion Energy ( LIFE ) are briefly introduced. It is based on recent progress of National Inertial Facility ( NIF ) and related research have begun. On the assumption of using laser energy of 1.4 to 2.0 MJ, the target fusion gain G=25~30, the repetition rate 10 to 15 Hz, the fusion power of 350 to 500 MW or neutron source power of 1.3×1020 to 1.8×1020 n/s could be achieved. For a sub-critical fission blanket driven by this fusion neutrons power, energy multiplication M of 4~10 and several GW of thermal power could be obtained. By novel design on fuel pins, burnup more than 90% would be achieved for heavy metals in the blanket. Inertial Confinement Fusion-fission energy is a promising concept, which characterized by inherent safety, richness in nuclear fuel resources, minimization of nuclear waste, non-CO2 emitting ,and it is a sustainable energy source.     

Calculations of the 3-D neutronics for CH HCSB TBM with 3×3 sub-modules

Using three dimension MCNP code and FENDL2.0 data library, neutronics calculation for a HCSB (helium cooling solid breeder) TBM (test blanket module) with 3×3 sub-modules has been performed. Local tritium breeding ratio (TBR) of 0.907, total tritium generation rate of 0.0175g·d-1, peak power density of 9.27MW·m-3 and total power deposit of 0.422MW·m-3 are obtained under neutron wall loading of 0.78MW·m-2 and duty factor of 22%.     

CFETR 堆体活化计算与分析

运用FLUKA计算程序对中国聚变工程实验堆(CFETR)进行了一维模拟活化运算,得出了产氚包层、屏蔽层、真空室结构材料、环向场线圈等模块的中子活化特性。计算结果表明,在聚变堆以200MW聚变功率持续稳态运行一年后,刚停堆时堆体的总活度为1.05×10 ¹⁹ Bq,停堆十年后堆体总活度为1.03×10 ¹⁷ Bq,此时堆体的主要残留放射性核素为⁵⁵ Fe。研究结果表明,目前CFETR的设计不存在突出的放射性环境安全问题。     

Calculations of the 3-D neutronics for CH HCSB TBM with 3×3 sub-modules

Using three dimension MCNP code and FENDL2.0 data library, neutronics calculation for a HCSB (helium cooling solid breeder ) TBM ( test blanket module ) with 3×3 sub-modules has been performed. Local tritium breeding ratio (TBR) of 0.907, total tritium generation rate of 0.0175 g•d^-1, peak power density of 9.27MW•m^-3  and total power deposit of 0.422MW•m^-3 are obtained under neutron wall loading of 0.78MW•m^-2 and duty factor of 22%.     

高增益包层冷却系统设计及初步热工水力分析北大核心CSCD

高增益包层氚增殖率能够达到1.5以上,能量放大倍数约为5,包层燃料区平均功率达50MW/m3,针对包层存在高功率密度区的这一特点,设计了采用迂回流动方案的水冷系统,主要由内嵌冷却管和汇总分流腔组成。建立了包括第一壁和燃料区的包层三维热工水力计算模型,利用CFD程序FLUENT对冷却系统进行模拟分析,研究了稳态工况条件下包层关键区域的整体热工水力特性。结果表明,该水冷系统流量分配合理,燃料区冷却剂压降为102kPa,出口温度为594K,符合设计预期。包层温度分布结果表明各区域最高温度均满足限值要求,冷却系统能够有效载出包层内裂变反应释放的热量。     

聚变裂变混合堆处理高放超铀废物的研究

提出一个燃烧高放超铀废物的思路,即在外部聚变中子源驱动下,把燃烧超铀锕系元素和钍铀燃料循环相结合.并且设计相应的一维模型,使用开发的燃耗计算程序ONESN_BURN和新制作的数据库对模型进行计算和分析.通过计算,得到锕系元素的放射性,生物潜在危害因子,高放超铀锕系废物的密度和非常深的燃耗深度等.比较聚变裂变混合堆与传统的热堆,发现中子能谱越硬,对燃烧超铀锕系元素越有效.     

Possible use of accelerator driven subcritical reactors for minor actinides incineration

The possible interest of accelerator driven subcritical reactors for minor actinides incineration is examined. The physics of neutron multiplying systems is recalled. The differences between critical and subcritical reactors' control are described, with emphasis on the importance of the delayed neutrons fraction. The minor actinides fuel evolution is studied with the conclusion that fast neutron spectra are clearly more efficient then thermal neutron spectra. It is, also, shown that characteristic times for incineration should be in the order of 10 years. The number of minor actinides incinerators necessary for 60 PWRs is estimated to be about 6 with total thermal power of 9 GW. These reactors will, also, be able to transmute essentially all ⁹⁹Tc and ¹²⁹I produced by the 60 PWR. The excess electricity cost for MA incineration is estimated to be about 5%.     

CFETR水冷陶瓷增殖剂包层中子学分析

基于中国聚变工程实验堆(CFETR)水冷陶瓷增殖剂(WCCB)三维中子学模型,应用蒙特卡罗输运程序MCNP5和IAEA聚变评价核数据库FENDL2.1,完成了WCCB中子学性能分析。研究了在200MW、500MW、1.0GW、1.5GW聚变功率下中子壁载荷(NWL)、氚增殖率(TBR)、核热沉积以及包层材料的辐照损伤。结果显示,目前WCCB包层核分析结果满足CFETR设计要求。     

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This paper considers the current China fusion engineering test reactor (CFETR) design, and simplifies it to a one-dimensional model. With the multi-particle transport code FLUKA, the neutron activation character of the tritium breeding blanket, shielding layer, vacuum vessel material and TFC of CFETR has been calculated to verify the radiation safety of the present design. The related results provide data reference for designing the components of CFETR and for further neutron activation analysis and calculation. The calculation results show that under the circumstances of one year operation with 200WM fusion power, the total radioactivity is 1.05×10 ¹⁹ Bq after shutdown and 1.03×10 ¹⁷ Bq after cooling for ten years. The primary residual nuclide is⁵⁵ Fe after decaying for ten years. It shows that there isn’t seriously activation safety issue.     

氘氚聚变中子发生器旋转氚靶传热特性研究

强流氘氚中子发生器可用于模拟聚变堆中子环境, 对于开展聚变堆包层材料相关实验研究具有重要意义. 本文提出了一种用于10¹² n·^-1量级氘氚中子发生器HINEG (high intensity neutron generator)的旋转氚靶系统设计方案, 并对其技术难点和强化传热方法进行了介绍. 为考查该氚靶系统的传热特性, 利用Computational Fluid Dynamics方法对冷却水层厚度、冷却水流速和氚靶系统旋转速度对靶面冷却的影响进行了分析, 并对不同热功率密度下靶面的传热过程进行了研究. 结果显示, 大的水层厚度、大的冷却水流速和高的靶系统旋转速度有利于靶面的冷却, 但水层厚度和水流速的变化对靶面传热影响较小. 一定条件下靶面所承受的热功率密度不能超过某个限值.     

次临界能源堆物理设计进展

聚变-裂变混合能源堆包括聚变中子源和次临界能源堆,主要目标是生产电能。回顾了国内外混合堆的发展历史,给出混合能源堆设计的边界条件和约束条件,说明次临界能源堆以铀锆合金为燃料、水为冷却剂的设计思想。利用输运燃耗耦合程序MCORGS计算了混合能源的燃耗,给出了中子有效增殖因数、能量放大倍数和氚增殖比等物理量随时间的变化。通过分析能谱和重要核素随燃耗时间的变化,说明混合能源堆与核燃料增殖、核废料嬗变混合堆的不同特点。论述了混合堆的热工设计并进行了安全分析。对于燃耗数值模拟程序,通过多家对算,保证其计算结果的可信性。针对次临界能源堆的特点,利用贫铀球壳建立了贫铀聚乙烯装置和贫铀LiH装置,并且专门设计加工了天然铀装置,开展铀裂变率、造钚率、产氚率等中子学积分实验,验证了数值模拟的可靠性。