产氚聚变堆概念设计系统研究

提出了基于球环类型的先进氚生产堆概念设计,它是聚变能发展的中间应用。与传统托卡马克氚生产堆不同,设计中利用了球形环的先进等离子体物理性能和紧凑的结构特征,并尽量利用真空室内的空间安置氚生产包层以减少氚泄露而增加氚增殖率,达到年生产氚1000 g的目标,相应的堆利用因子为40％。在2D中子学计算的基础上提出了较为完整的初步概念设计。逐项进行了分析,同时对设计的风险、不确定性和后备方案也做了概括的解释。为下一步更详细、具体的概念设计提供了直接的依据和重要的参考价值。     

Effect of Nuclear Data Libraries on Tritium Breeding in a (D–T) Fusion Driven Reactor

In design a Deuterium–Tritium (D–T) fusion driven hybrid reactor, neutronics and nuclear data libraries have an essential role for reliable neutronics calculations. Therefore, nuclear data libraries are very important to calculate of the neutronic parameters and selection of tritium breeder materials to be used in the blanket. In this study tritium breeding performances of candidate tritium breeding materials, namely, Li₂O, LiH, Li₂TiO₃, Li₂ZrO₃ and Li₄SiO₄ in a (D–T) driven fusion–fission (hybrid) reactor is investigated based on three dimensional (3-D) and one dimensional (1-D) neutronic calculations. 3-D and 1-D neutron transport calculations are performed with Monte Carlo transport code (MCNP 4C), SCALE 5 and ANISN nuclear data codes to determine the tritium breeding ratio (TBR) of the blanket. The effects of different nuclear data libraries on TBR are examined and TBR calculation results are comparatively investigated.     

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.     

Neutronic Comparison of Tritium－Breeding Performance of Candidate Tritium－Breeding Materials

Tritium self-sustainment, which will meet the fuel requirement of fusion reactor, is one of the key issues of fusion power development. The tritium breeding performances of various tritium-breeding materials are compared based on a series of neutronics calculations using three-dimensional Monte Carlo neutron-photon transport code MCNP/4C with the IAEA FENDL-2 data library. The effects of the dimensions of the tritium-breeding zone and the enrichment of 6Li on Tritium Breeding Ratio (TBR) are analyzed. The effects of Be as a neutron multiplier on TBR are also calculated.     

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.     

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。     

球环型产氚聚变堆概念设计研究

该研究基于球环类型的先进氚生产堆概念而设计,是聚变能发展的中间应用。与传统托卡马克氚生产堆不同,该设计利用球形环的先进等离子体物理性能和紧凑的结构特征,尽量利用真空室内的空间安置氚生产包层以减少氚泄露而增加氚增殖率,相应的堆利用因子为 40%。在二维中子学计算的基础上提出了较为完整的初步概念设计。在逐项分析的基础上对设计的风险、不确定性和后备方案也做了概括的解释。为下一步更详细具体的概念设计提供了直接的依据,具有重要的参考价值。     

A neutron poison tritium breeding controller applied to a water cooled fusion reactor model

The generation of tritium in sufficient quantities is an absolute requirement for a next step fusion device such as DEMO due to the scarcity of tritium sources. Although the production of sufficient quantities of tritium will be one of the main challenges for DEMO, within an energy economy featuring several fusion power plants the active control of tritium production may be required in order to manage surplus tritium inventories at power plant sites. The primary reason for controlling the tritium inventory in such an economy would therefore be to minimise the risk and storage costs associated with large quantities of surplus tritium. In order to ensure that enough tritium will be produced in a reactor which contains a solid tritium breeder, over the reactor's lifetime, the tritium breeding rate at the beginning of its lifetime is relatively high and reduces over time. This causes a large surplus tritium inventory to build up until approximately halfway through the lifetime of the blanket, when the inventory begins to decrease. This surplus tritium inventory could exceed several tens of kilograms of tritium, impacting on possible safety and licensing conditions that may exist.This paper describes a possible solution to the surplus tritium inventory problem that involves neutron poison injection into the coolant, which is managed with a tritium breeding controller. A simple PID controller and is used to manage the injection of the neutron absorbing compounds into the water coolant of a stratified blanket model, depending on the difference between the required tritium excess inventory and the measured tritium excess inventory. The compounds effectively reduce the amount of low energy neutrons available to react with lithium compounds, thus reducing the tritium breeding ratio. This controller reduces the amount of tritium being produced at the start of the reactor's lifetime and increases the rate of tritium production towards the end of its lifetime. Thus, a relatively stable tritium production level may be maintained, allowing the control system to minimize the stored tritium with obvious safety benefits. The FATI code (Fusion Activation and Transport Interface) will be used to perform the tritium breeding and controller calculations.     

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.     

Tritium handling, breeding, and containment in two conceptual fusion reactor designs: UWMAK-II and UWMAK-III

Tritium is an essential component of near-term controlled thermonuclear reactor systems. Since tritium is not likely to be available on a large scale at a modest cost, fusion reactor designs must incorporate blanket systems which will be capable of breeding tritium. Because of the radiological activity and capability of assimilation into living tissues, tritium release to the environment must be strictly controlled. The University of Wisconsin has completed three conceptual designs of fusion reactors, UWMAK-I, UWMAK-II, and UWMAK-III. This report discusses the tritium systems for UWMAK-II, a reactor design with a helium cooled solid breeder blanket, and UWMAK-III, a reactor design with a high-temperature liquid breeder blanket. Tritium systems for fueling and recycling, breeding and recovery, and plant containment and control are discussed.     

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。     

On the Tritium Breeding Capability of Flibe, Flinabe, and Li20Sn80 in a Fusion-Fission (Hybrid) Reactor

In a commercial (DT) driven fusion reactor, the tritium breeding ratio per incident fusion neutron must be greater than 1.05 to maintain tritium self-sufficiency for the driver. In this study tritium breeding capability of three different coolants, namely Flibe (LiF·BeF₂), Flinabe (LiF·NaF·BeF₂), and Li₂₀Sn₈₀ in a (DT) driven fusion-fission (hybrid) reactor was investigated for different refractory alloys (W-5Re, TZM, T111, and Nb-1Zr) as structural material. Neutron transport calculations were conducted with the help of SCALE 4.3 SYSTEM by solving the Boltzmann transport equation with code XSDRNPM. The contribution of Flibe, Flinabe, and Li₂₀Sn₈₀ with respect to ⁶Li enrichment in their lithium content to overall TBR was investigated. In addition, the effect of structural material type on TBR was examined.     

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.     

Proposal of an MFE-IFE cooperative system for a steady-state tritium balance

A new concept of a fusion reactor system, MFE-IFE cooperative system, is proposed. This concept combines the merits of a small-size MFE reactor and a dry-wall IFE reactor and aims at sufficient amount of tritium production and electricity generation without advanced technology. Design window analysis shows a NIF-scale (5 m chamber radius) dry-wall laser fusion reactor with a ～1 GWth fusion output and net tritium breeding ratio (TBR) of 1.74 can sustain an MFE power plant with a fusion power of 3 GWth and net TBR of 0.96. Although more detailed quantitative analyses are required, this concept can be a possible solution for a simultaneous achievement of tritium self-sufficiency and significant net electricity generation.     

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.     

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。     

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.     

Impact of Solid Breeder Materials on Tritium Breeding in a Hybrid Reactor

Tritium breeding ratio (TBR) is one of the important parameters in design of a Deuterium–Tritium (DT) driven hybrid reactor. Therefore, selection of tritium breeder materials to be used in the blanket is very crucial. In this study, tritium breeding potential of the solid breeders, namely, or in a (DT) fusion driven hybrid reactor fuelled with or was investigated. For this purpose in addition to these solid breeders, different types of liquid breeders, namely natural lithium, Flibe, Flinabe and were used to examine the tritium breeding behavior of liquid–solid breeder couple combinations. Numerical calculations were carried out by using Scale 4.3. According to numerical results, the blanket with fuel using natural lithium as coolant and as solid breeder had the highest TBR value.     

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.     

Nuclear Performance of Two-Phase Flow Cooling D-T Fusion Reactor Blankets

The aim of the present paper is to analyze the nuclear performance of a typical D-T fusion reactor blanket cooled by two-phase flow, and, in particular, the dependence of tritium breeding ratio (TBR), nuclear heating and neutron (energy) leakage on design variations such as the volume fraction γ occupied by coolant materials.

The value of γ plays a central role in determining the nuclear performance of the blanket considered. The TBR and nuclear heating decrease with decreasing γ while the inverse trend is found for the leakage from the blanket. To obtain the TBR greater than unity would require γ at least 30%. The feasibility of the two-phase flow cooling concept for D-T reactor blankets is contigent upon finding the way of taking advantage of the many good features associated with the flow, even at such γ.