Neutronics study for the installation of the LBVM in the medium flux area of IFMIF

IFMIF (International Fusion Materials Irradiation Facility) will be a fusion dedicated facility producing a large amount of neutrons with the appropriate energy spectrum to test materials and subcomponents for DEMO and future Fusion Power Plants.While the high flux area of IFMIF will be devoted to reduced activation structural materials for first wall and blanket, the medium flux area will be dedicated to functional materials for breeder blankets. In particular, the Liquid Breeder Validation Module (LBVM), will host experiments related with functional materials for liquid breeder blankets. Since IFMIF neutron spectra have been intended to fit the most irradiated areas of a fusion reactor in the high flux area, the irradiation conditions in the LBVM placed in the medium flux area of IFMIF have been assessed. The effect of some neutron shifter/reflector components to optimize the neutron spectra have been evaluated in order to find out the proper irradiation conditions for functional materials for liquid breeder blankets.Therefore, the objective of this report is to summarize the neutronic calculations developed to evaluate the viability of IFMIF neutron source to perform relevant irradiation experiments on functional materials for liquid breeder blanket concept for future nuclear fusion power reactors (ITER, DEMO). The irradiation parameters evaluated for this purpose are: the tritium production for liquid breeder material (Pb–17Li) and the damage dose (dpa) and gas production to damage dose ratios for Al₂O₃ and SiC functional materials.The main conclusion is that, it is possible to perform relevant irradiation experiments on functional materials for liquid breeder blanket concept for the future nuclear fusion reactor DEMO. Nevertless, the use of some shifter components will be needed to optimize some irradiation parameters.     

Parametric neutronic analysis of HCLL blanket for DEMO fusion reactor utilizing vacuum vessel ITER FDR design

Three-dimensional parametric neutronics calculations using the Monte Carlo code MCNP-4C have been performed for a DEMO-type reactor based on the Helium-Cooled Lithium-Lead (HCLL) blanket. The aim of the analysis was to minimize the radial blanket thickness, while ensuring tritium self-sufficiency and to assess the shielding performance of the reactor in terms of the radiation loads to the super-conducting toroidal field (TF) coils. It was found that tritium self-sufficiency can be achieved with a breeder zone thickness reduced to no more than 55 cm at a ⁶Li enrichment of 90%. Assuming a ⁶Li enrichment of 60%, a breeder zone thickness of 60 cm is required to achieve the target TBR of 1.10 which is assumed to be sufficient to cover potential tritium losses and uncertainties. With regard to the shielding performance it was found that the design limits for the radiation loads to the TF-coil can be met with radial blanket thicknesses of 75 cm, 60 cm and 55 cm utilizing a two-component shield of Eurofer steel and tungsten carbide between the breeder zone and the vacuum vessel. The blanket variants with larger radial breeder zone show better shielding performances due to the reduced Eurofer shielding material acting as gamma radiation emitter in between the breeder zone and the vacuum vessel. In particular the radiation dose absorbed in the Epoxy insulator was shown to be the most critical quantity in this regard.     

Lead-lithium eutectic material database for nuclear fusion technology

Fully validated material databases are needed for coherent technological developments in any R&D field. For nuclear fusion technology (NFT), within a near-term perspective of qualification and licensing of nuclear components and systems, this goal is both compulsory and urgent. This mandatory requirement applies for the particular case of the Pb-Li eutectic database as fusion reactor material. Pb16Li is today a reference breeder material in diverse fusion R&D programs worldwide. Technical consensus on most part of the material database inputs seems a major technological objective. In this work Pb16Li material database inputs for NFT have been systematically reviewed. Database inputs (bulk, thermal, physical-chemistry properties, and H-isotopes transport) are discussed and extended to base magnetohydrodynamic (MHD) properties, values for non-dimensional parameters and pipe/channel correlations in 2-phases dispersion models. Ongoing efforts to develop the Pb16Li material database as a computing expert system are reported.     

单斜相偏锆酸锂陶瓷氚增殖材料研究

单斜相偏锆酸锂是聚变堆中最有前途的氚增殖材料之一。该材料具有较高的锂原子密度和优异的氚释放行为，成为近年来最受重视的陶瓷氚增殖材料。在Ｌｉ２Ｏ－ＺｒＯ２二元系中，存在九种不同的锆酸锂相，因此制备单一相的锆锂陶瓷十分困难。本文作者以干法制备工艺为基础，从热扩散的角度分析了Ｌｉ２ＺｒＯ３的形成机理，改进了制备工艺，成功地制备出了单斜相Ｌｉ２ＺｒＯ２陶瓷材料粉末样品。     

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。     

Lithium and tritium diffusion in lithium oxide (Li2O), a molecular dynamics simulation

Molecular dynamics simulations of Li₂O were performed in the microcanonical ensemble at several different temperatures in order to study the lithium diffusion process, and a preliminary exploration of the diffusion of tritium was performed. Different Li/O molar ratios were used to investigate the role of non-stoichiometry in the Li diffusion processes. The mechanism of lithium diffusion as a function of temperature is proposed based on the analysis of our simulations and a model is proposed to explain the overall behaviour of the lithium diffusion coefficient as a function of temperature. Our simulations suggest what is the role of hydrogen in the tritium release from breeder ceramic materials.     

3D thermo-fluid MHD simulation of single straight channel flow in LLCB TBM

India is developing lead lithium cooled ceramic breeder (LLCB) blanket for its DEMO fusion reactor. The mock-up blanket (TBM), using this concept, will be tested in ITER for its tritium breeding and high-grade heat extraction efficiency. In this TBM, pressurized helium is used to remove the heat from first wall, top and bottom plates of TBM. The Pb–Li is used to extract heat from the breeder zones. The flow of Pb–Li with average velocity 0.1 m/s inside the channel can be significantly modified due to MHD effects, which arise because of the presence of strong toroidal magnetic field. A numerical approach is established to capture this flow modification at higher Hartmann numbers (≥20,000). As a validation part of the developed code, MHD phenomenon is studied in 2-D square geometry and numerically obtained velocity profile is compared with available Hunt's analytical results. Thermo-fluid MHD analysis using this code, has been carried out for single rectangular duct of LLCB TBM. The heat transfer has been studied by keeping hot breeders at both sides of the flow channel. The results suggest modification in steady state MHD velocity profile as the liquid flows along the flow length. However, the temperature in various zone remains well within the maximum allowable limit.     

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。     

Solid breeder thermal transient under purge-line-break accidents

The breeder thermal performances under a purge line break have been analyzed for two blanket design options: a blanket design using a packed breeder bed and a blanket design using a sintered breeder product. Under a purge line break open to a vacuum environment, the packed bed breeder temperature exceeds its operating temperature limit at a faster rate than that of the sintered breeder blanket design for the same breeder temperature gradient. Depending on the breeder material and nominal operating conditions, the breeder reaches its maximum operating temperature in time ranging from 32 seconds to 125 seconds for a break area of 10 cm² in packed bed designs. However for the sintered product design, the consequence of this transient might not result in the breeder exceeding its maximum operating temperature if a reasonable contact pressure could be established at the interface. To reduce the safety hazards, the tritium concentration build up in the vacuum vessel in conjunction with the purge gas pressure inside the blanket module should be used as a measure for initiating the reactor shutdown for this type of accident. The consequence of the purge line break outside the vacuum vessel on the breeder transient thermal performance is less significant because of a longer transient time involved.     

The solubility of aluminium in Pb-17Li

The saturation solubility of aluminium in Pb-17Li has been measured over the temperature range envisaged for a Pb-17Li tritium breeder/coolant blanket for use in a fusion reactor. The solubility is given by the equation log¹⁰S(wppm) = 6.249 – 2784.9/T(K) for T = 525 – 813 K.The results are compared to literature values for the solubility of aluminium in pure lead and show good agreement. A value for the enthalpy of solution of + 55.8 kJ mol^-1 has been calculated.     

Comparative studies for two different orientations of pebble bed in an HCCB blanket

The Indian Test Blanket Module(TBM) program in ITER is one of the major steps in its fusion reactor program towards DEMO and the future fusion power reactor vision. Research and development(RD) is focused on two types of breeding blanket concepts: lead–lithium ceramic breeder(LLCB) and helium-cooled ceramic breeder(HCCB) blanket systems for the DEMO reactor. As part of the ITER-TBM program, the LLCB concept will be tested in one-half of ITER port no. 2, whose materials and technologies will be tested during ITER operation. The HCCB concept is a variant of the solid breeder blanket, which is presently part of our domestic RD program for DEMO relevant technology development. In the HCCB concept Li_2TiO_3 and beryllium are used as the tritium breeder and neutron multiplier, respectively, in the form of a packed bed having edge-on configuration with reduced activation ferritic martensitic steel as the structural material. In this paper two design schemes, mainly two different orientations of pebble beds, are discussed. In the current concept(case-1), the ceramic breeder beds are kept horizontal in the toroidal–radial direction. Due to gravity, the pebbles may settle down at the bottom and create a finite gap between the pebbles and the top cooling plate, which will affect the heat transfer between them. In the alternate design concept(case-2), the pebble bed is vertically(poloidal–radial) orientated where the side plates act as cooling plates instead of top and bottom plates. These two design variants are analyzed analytically and 2 D thermal-hydraulic simulation studies are carried out with ANSYS, using the heat loads obtained from neutronic calculations.Based on the analysis the performance is compared and details of the thermal and radiative heat transfer studies are also discussed in this paper.     

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.     

Options and methods for instrumentation of Test Blanket Systems for experiment control and scientific mission

Europe is currently developing two reference breeder blankets concepts for DEMO reactor specifications that will be tested in ITER under the form of Test Blanket Modules (TBMs): the Helium-Cooled Lithium-Lead (HCLL) concept which uses the eutectic Pb-16Li as both breeder and neutron multiplier; the Helium-Cooled Pebble-Bed (HCPB) concept which features lithiated ceramic pebbles as breeder and beryllium pebbles as neutron multiplier. Each TBM is associated with several sub-systems required for their operation; together they form the Test Blanket System (TBS). This paper presents the state of HCLL and HCPB TBS instrumentation design. The discussion is based on the systems functional analysis, from which three main categories of instrumentation are defined: those relevant to safety functions; those relevant to interlock functions; those designed for the control and scientific exploitation of the devices based on the TBM program objectives.     

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.     

R&D status on Water Cooled Ceramic Breeder Blanket Technology

Japan Atomic Energy Agency (JAEA) is performing the development of a Water Cooled Ceramic Breeder (WCCB) Test Blanket Module (TBM) as one of the most important steps toward DEMO blanket. Regarding the blanket module fabrication technology development using F82H, the fabrication of a real scale mockup of the back wall of TBM was completed. In the design activity of the TBM, electromagnetic analysis under plasma disruption events and thermo-mechanical analysis under steady state and transient state of tokamak operation have been performed and showed bright prospect toward design justification. Regarding the development of advanced breeder and multiplier pebbles for DEMO blanket, fabrication technology development of Li rich Li₂TiO₃ pebble and BeTi pebble was performed. Regarding the research activity on the evaluation of tritium generation performance, the evaluation of tritium production and recovery test using D-T neutron in the Fusion Neutronics Source (FNS) facility has been performed. This paper overviews the recent achievements of the development of the WCCB Blanket in JAEA.     

Study on the Chemical Compatibility Study Between Li<Subscript>2</Subscript>TiO<Subscript>3</Subscript> Pebbles and 14Cr-ODS Steel

In order to investigate the chemical compatibility between tritium breeder Li₂TiO₃ pebbles and tritium breeder blanket material oxide dispersion strengthened (ODS) steel, the contact interface between Li₂TiO₃ pebbles and ODS steel heated in argon atmosphere at 500, 600 and 700 °C for 300 h was studied. It was found that the ions of pebbles could diffuse and corrode with the cladding material after a long-time reaction at high temperature. The corrosion area formed on the surface of Li₂TiO₃ pebbles was small. With the increase of temperature, a zone with enriched iron was found on the surface of the pebble. This part of the surface was the direct contact surface between the pebble and the steel. At the same time, the relative density of the pebbles increased and the crush load was decreased to 30 N. In addition, a slight corrosion phenomenon was found on the surface of ODS steel. It has been proved that the main components of the corrosion products were the complex oxide containing Fe and Cr and the complex oxide containing Li and Fe.     

Development of the water cooled lithium lead blanket for DEMO

The water cooled lithium lead (WCLL) blanket, based on near-future technology requiring small extrapolation from present-day knowledge both on physical and technological aspect, is one of the breeding blanket concepts considered as possible candidates for the EU DEMOnstration power plant.In 2012, the EFDA agency issued new specifications for DEMO: this paper describes the work performed to adapt the WCLL blanket design to those specifications.Relatively small modules with straight surfaces are attached to a common Back Supporting Structure housing feeding pipes. Each module features reduced activation ferritic-martensitic steel as structural material, liquid Lithium-Lead as breeder, neutron multiplier and carrier. Water at typical Pressurized Water Reactors (PWR) conditions is chosen as coolant.A preliminary design of the equatorial outboard module has been achieved. Finite elements analyses have been carried out in order to assess the module thermal behavior. Two First Wall (FW) concepts have been proposed, one favoring the thermal efficiency, the other favoring the manufacturability. The Breeding Zone has been designed with C-shaped Double-Walled Tubes in order to minimize the Water/Pb-15.7Li interaction likelihood.The priorities for further development of the WCLL blanket concept are identified in the paper.