Tritium breeding control within liquid metal blankets

A key requirement for DEMO is the on-site breeding of tritium. In order to do this, a robust control system must be employed to ensure enough tritium is being bred to sustain the fusion reactor, whilst not breeding an amount which would exceed the plant's tritium inventory license. A tritium breeding method which is cost effective and reduces radioactive waste for disposal is that of the liquid metal breeder such as those based around LiPB and FLiBe. This paper focuses on the modeling of a simplified fusion reactor design with a LiPb blanket with linked radiation transport, nuclide burn-up and control theory. Two simple models were simulated using the FATI code which incorporated a PID (proportional integral derivative) controller that adjusted the Li6/Li7 ratio in order to increase/decrease tritium production based on the difference between the measured excess tritium inventory and the desired excess inventory. The modelling has initially demonstrated that a linear PID controller has the capability to manage tritium production within a LiPb liquid blanket.     

氟盐冷却高温堆氚输运特性数值研究

氚的控制是限制氟盐冷却高温堆（FHR）发展的关键问题,欲实现氚的有效控制,首先需明确氚在熔盐堆一回路中的输运行为。本文阐明了氚在熔盐堆一回路中的输运特性,包括氚的产生及存在形态的分化、石墨对氚的吸附、氚在熔盐中的溶解与扩散以及氚在管壁材料中的渗透等。针对氚在熔盐堆一回路中的输运行为,建立了数学物理模型,基于FORTRAN语言开发了适用于FHR的氚输运特性分析程序TAPAS。通过将实验数据与程序计算结果对比,说明了TAPAS程序计算的合理性和准确性。利用TAPAS对模块化移动式氟盐冷却高温堆（TFHR）中氚的输运特性进行了分析。计算表明,TFHR的初始产氚率约为5.54×10^-8 mol/s,一回路中的氚主要以T₂形式存在,腐蚀反应主要发生在热管段入口处。反应堆运行25 EFPD（等效满功率天）后,石墨吸附氚达到限值。反应堆稳态运行时,T₂向管壁表面的渗透速率可视为常数,其值为8.35 μmol/EFPD。本研究可为FHR的研究设计和辐射防护提供参考。     

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.     

Experimental design of tritium extraction loop from lead lithium eutectic

Effective tritium breeding achievable in Test Blanket Module (TBM) is a major issue for sustainable fusion energy program. Equally important is tritium extraction to recover and recycle tritium back to fusion reactor. Tritium extraction from lead lithium is much more complicated than from purge gas due to low tritium extraction efficiency in transfer step to gas phase and the limitations imposed on space and lead lithium inventory in port cell. Earlier investigations do suggest the preference of packed columns over bubble columns. Theoretical models based on axial dispersion plug flow in liquid and gas proposed for bubble columns and packed columns are reinvestigated for different boundary conditions.This paper highlights the critical issues of experimental design based on tritium extraction efficiency and its impact on recovery loop. Steady state closed loop for absorption and stripping of hydrogen isotopes using inert gas is designed along with the associated auxiliaries.     

液态锂锡合金氚增殖行为的理论分析

采用气-液两相界面模型和与时间有关的扩散理论及本征函数展开的方法,模拟了Li25Sn75合金的氚增殖行为.计算结果表明:在14 MeV能量下,天然Sn的(n,2n)反应宏观截面相对较小,只有1.5 b;7Li、6Li产氚随时间变化的规律与LiPb合金、Li2O介质是一致的;Li25Sn75合金对模型厚度比较敏感,随着厚度和6Li丰度的增加,氚增殖比(Tritium Breeding Ratio,TBR)保持上升的趋势.     

氚/单晶硅器件辐伏电池模型及样机的长期稳定性研究

采用氚化钛源原位辐照和加速器低能电子束加速辐照实验,研究基于氚化钛/单晶硅PN结器件的氚辐伏电池模型和实验室组阵型电池原型样机的长期稳定性。测试电池模型和电池样机的氚辐伏输出随辐照时间的变化,分析辐照对单晶硅PN结型器件的本征暗特性和器件表面层材料缺陷的影响。结果表明,氚化钛源原位辐照电池模型在115 d的辐照中辐伏输出没有明显的衰减,辐照后单晶硅器件的本征暗特性曲线变化微小。电池模型的加速器低能电子束加速辐照实验表明,加速辐照在相同电子注量下对电池造成远大于氚源原位辐照的性能损伤,但损伤仅在辐照最初期快速产生,随后基本保持稳定,电子顺磁能谱（ESR）测试加速辐照60 min单晶硅器件材料的缺陷没有明显增加。组阵型实验室电池原型样机在64个月的室温储存中,基本单元的辐伏输出性能衰减比氚的自发衰变衰减有小幅增大,增大幅度小于11.4%;另外,组阵中单元之间串并联电连接有部分失效,这是后续应重点关注的问题。     

CFETR氚自持分析评估与验证策略

中国聚变工程实验堆(China Fusion Engineering Test Reactor,简称CFETR)的主要目标之一是实现氚自持.采用氚平衡法对CFETR不同运行工况下的氚自持条件进行了分析评估.结果表明:在500 MW运行阶段,CFETR实现氚自持所需的最小氚增殖比(TBRr)为1.098,小于CFETR增...     

IBIL在氚增殖材料研究中的应用

在氚增殖材料辐照效应的研究中,离子激发发光(ion beam induced luminescence,IBIL)是一种高效实用的实时分析技术。本文介绍了国外MeV离子束对多种氚增殖材料的IBIL研究。研究表征了样品中的辐照缺陷特征及其演变情况,对辐照缺陷的产生机制进行了讨论,并提出辐照过程中的相关动力学模型。最后,介绍了北京师范大学串列加速器上IBIL装置应用现状,并对IBIL应用在氚增殖材料研究中的前景进行了讨论。     

内标淬灭校正法测定尿中氚浓度

氚引起的内照射对人体伤害非常大,因此评估涉氚人员的氚内照射水平十分重要。为满足快速准确评估人员氚摄入水平的要求,本研究在原尿直接测量法的基础上,采用内标淬灭校正法,对三名涉氚工作人员尿液中的氚放射性浓度进行测量,评定测量结果的不确定度,并估算该方法的探测下限。测量结果表明,李、胡、黄三位工人的尿中氚的放射性浓度分别为94.9、554、898 Bq/L,与氧化蒸馏法测量结果的偏差均在4%以内。该方法操作简便、快捷,准确性较高,可以满足涉氚工作人员的尿氚监测和内照射剂量快速准确评价的要求。     

Nuclear design for a heavy ion-beam fusion reactor,HIBLIC

A nuclear analysis was carried out for a heavy ion-beam fusion reactor, HIBLIC. The analysis includes the target and chamber neutronics, time-dependent radiation damage in the first wall, and radiation streaming through beam ports. It is found that the reactor chamber is characterized by its high tritium breeding ratio, low radiation damage in the second wall, and low induced activity. To reduce the radiation damage in the superconducting, focusing magnets, tapering the beam ports along the direct line-of-sight component of the source neutron is necessary in the magnet regions and also in the collimator region.     

Effect of outside tritium source on tritium balance of a D-T fusion reactor

Attainable tritium breeding ration in the blanket system must be larger than the required breeding ratio when no effective tritium resources from outside are expected. It is revealed recently that a considerable amount of tritium can be trapped to the re-deposition layer of the first wall materials and that the time constant of this phenomenon is rather long. Then, the tritium breeding ratio around 1.1 is required in the blanket system when 3 years is claimed for the tritium doubling time to prepare tritium for the initial inventory of a next reactor. Construction of an outside tritium supply is one of the possible ways to compensate the lack of tritium because it is generally considered that the attainable tritium breeding ratio in the solid breeder system is around 1.05. It is reported recently that a high-temperature gas-cooled reactor can produce 10 kg of tritium per year. The preferable amount of tritium production rate of the outer tritium supply is discussed in this study from the viewpoint of tritium balance in a D-T power reactor.     

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.     

Alternate approach to inertial confinement fusion with low tritium inventories and high power densities

A low-tritium-inventory, high-power-density, pool-type chamber approach to inertial confinement fusion is introduced. The concept uses target designs with internal tritium and³He breeding, eliminating the need for a lithium-breeding blanket. The fraction of the fusion energy carried out by neutrons is estimated as 10%, compared with 70% in a typical D-T system, and the neutron spectrum is softer. Liquid metals other than lithium that are less chemically reactive, such as lead, can be used for first-wall protection. The reduced neutron component and the elimination of the need for a thick lithium blanket for tritium breeding lead to higher power densities and more compact chamber designs. The radiation damage at the first structural wall is reduced, leading to potentially longer wall lifetimes. A significant environmental advantage in terms of reduced radioactive release risks under operational and accident conditions is identified, primarily due to the one to two orders of magnitude reduction in the tritium inventories compared with D-T-based systems.     

HTR-10一回路冷却剂中氚活度的测量

详细介绍了测量10 MW高温气冷试验堆一回路冷却剂中氚活度的方法。设计适用于HTR-10特点的氚收集装置,先后两次收集冷却剂中的氚,制成液样进而用液闪法进行测量,并根据试验结果推算HTR-10一回路冷却剂中氚的总活度。针对两次试验结果进行分析并与理论计算值相比较,验证了理论计算的正确性并由此进一步证明高温气冷堆的燃料包覆颗粒对放射性产物的阻挡作用完好,反应堆对环境的氚释放完全在设计要求范围内,符合相应的国家标准。     

Neutronics calculations for normal conducting copper RiggatronTM tokamaks

Tritium breeding is found to be achievable for water cooled normally conducting copper magnet tokamaks that would be placed inside a cell containing the tritium breeding blanket.     

Effect of Lithium Enrichment on the Tritium Breeding Characteristics of Various Breeders in a Fusion Driven Hybrid Reactor

Selection of lithium containing materials is very important in the design of a deuterium–tritium (DT) fusion driven hybrid reactor in order to supply its tritium self-sufficiency. Tritium, an artificial isotope of hydrogen, can be produced in the blanket by using the neutron capture reactions of lithium in the coolants and/or blanket materials which consist of lithium. This study presents the effect of lithium-6 enrichment in the coolant of the reactor on the tritium breeding of the hybrid blanket. Various liquid–solid breeder couples were investigated to determine the effective breeders. Numerical results pointed out that the tritium production increased with increasing lithium-6 enrichment for all cases.     

氦-3回路中氚的辐射安全分析

氚安全是确保燃料元件堆内功率瞬态试验的关键因素之一。本文首先分析了氚的来源和危害,提出了氦-3回路氚的防护和去污措施,然后讨论了氚在正常运行和事故时释放到包容箱和工艺间的量和处理措施,最后评价了氦-3系统发生不同安全措施失效的事故情况下工作人员的氚内照射剂量。结果表明:系统正常运行时工作人员所受最大剂量为1.27 μSv/d,除了氚安全措施全部同时失效且HT短时间全部被氧化成HTO的极限事故以外,在一般事故情况下氚对工作人员产生的最大剂量小于10 mSv。     

压水堆主冷却剂中氚源项计算分析

压水堆主回路冷却剂流经堆芯时,水中固有及特加核素受中子辐照后会产生氚,氚几乎全部以气体和液体的形式排入环境,造成氚污染。因此,氚是压水堆辐射环境影响评价的主要关注内容之一。本文以AP1000为例,根据压水堆主回路冷却剂中氚的产生途径及其随时间的变化情况建立详细的计算模型,计算压水堆主回路冷却剂中的氚活度并分析各产氚途径对氚产生量的贡献。计算结果表明：主回路冷却剂中的氚主要来源于可溶性硼的中子活化和铀裂变,对氚产生量的贡献达80%以上;在⁷Li纯度为99.9%时,AP1000主回路中的年产氚量为5.23×10¹³ Bq,锂产氚量占总量的14.01%,随⁷Li纯度的增加,锂产氚量的贡献呈线性减小,在⁷Li纯度为99.99%时,锂产氚量占总量的3.18%。其他途径对氚的产生量贡献很小,可忽略。根据以上结果,可通过控制主回路冷却剂中添加的初始硼浓度、提高燃料包壳质量、增加LiOH中⁷Li的纯度等多种途径来降低主冷却剂中氚的产生量,从而减少氚对环境的放射性污染。     

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。     

聚变实验增殖堆氚系统设计研究(英文)

介绍了核工业西南物理研究院聚变实验增殖堆工程概要设计(FEB-E)中的氚系统设计研究。第一部分介绍包层氚增殖区的划分、几何尺寸、装料特征和用蒙特卡罗程序计算得到的液态锂中的氚浓度分布;第二部分描述根据聚变堆氚物理基础构造的氚循环系统,共分成 10 个子系统及它们之间氚的流程图。运用研制的程序SWITRIM 计算了各个子系统中的氚投料量随时间的变化,满功率运行一年后各个子系统中的氚投料量。研究结果表明起动 143 MW 聚变功率 FEB-E 堆所需要的初始氚投料量大约为 319 g。第三部分对不同的运行状态下的氚泄漏问题进行了分析。潜在的氚泄漏危险可能来自于偏滤器系统从等离子体中抽出的气体。得到的结论是提高FEB-E 堆芯等离子体的燃耗份额从而减少氚的通过量对降低氚的泄漏危险是重要的。