Multiplier Material Research Articles

Blistering Behavior of Beryllium and Beryllium Alloy under High-Dose Helium Ion Irradiation.

Beryllium (Be) has been selected as the solid neutron multiplier material for a tritium breeding blanket module in ITER, which is also the primary option of the Chinese TBM program. But the irradiation swelling of beryllium is severe under high temperature, high irradiation damage and high doses of transmutation-induced helium. Advanced neutron multipliers with high stability at high temperature are desired for the demonstration power plant (DEMO) reactors and the China Fusion Engineering Test Reactor (CFETR). Beryllium alloys mainly composed of Be12M (M is W or Ti) phase were fabricated by HIP, which has a high melting point and high beryllium content. Beryllium and beryllide (Be12Ti and Be12W) samples were irradiated by helium ion with 30 keV and 1 × 1018 cm-2 at RT. The microstructures of Be, Be12Ti and Be12W samples were analyzed by SEM and TEM before and after ion irradiation. The average size of the first blistering on the surface of Be-W alloy is about 0.8 μm, and that of secondary blistering is about 79 nm. The surface blistering rates of the beryllium and beryllide samples were also compared. These results may provide a preliminary experimental basis for evaluating the irradiation swelling resistance of beryllium alloy.

Investigation of hydrogen generation from the reaction of beryllium pebbles with water vapor for WCCB-TBM

Be pebble, as a neutron multiplier material, is being explored for the development of a Water-Cooled Ceramic Breeder Test Blanket Module (WCCB-TBM) in Japan for the International Thermonuclear Experiment Reactor (ITER) project. The Be pebbles were fabricated by the rotating electrode process (REP). However, as the safety of the test blanket module (TBM) could be compromised by hydrogen generation via Be–H2O reaction owing to the leakage of pressurized water from the cooling pipes inside TBM, the safety assessment of TBM must be conducted using the Be–H2O reactivity data for the Be pebbles whose fabrication method is the same as that for TBM (i.e., REP). Thus, in this study, a Be–H2O reactivity test was conducted using Be pebbles that were fabricated by REP to characterize the generated hydrogen and evaluate the hydrogen generation (HG) rates toward WCCB-TBM safety assessment. First, the Be–H2O reactivity data were obtained at ≥ 300 °C (the coolant temperature for the TBM). The experimental results and oxidation rate of the Be pebbles revealed that only the Be surface may have been oxidized at 300 °C–600 °C, although its inside was oxidized at 700 °C–1000 °C. Further, the experimental data–derived HG rates at 300 °C–1000 °C increased with the temperature and the tendency changed in the three temperature regions. The HG rates did not contradict those reported by previous studies. Additionally, these regions corresponded to each expected Be–H2O reaction behavior at 300 °C–1000 °C. At a Be temperature of < 600 °C, the generated hydrogen from TBM will not exceed 2.5 kg of a limitation by the water leakage on the estimation based on the HG rates and the qualitative effects of the safety functions.

Helium swelling behavior for neutron multipliers after irradiation with He ions at high temperatures

Beryllides as beryllium intermetallic compounds are the advanced multiplier materials as candidate materials for fusion demonstration power plant (DEMO) due to having good properties such as low swelling rate and stable chemical properties.Under DEMO operation environment, the generated He amounts in beryllium are estimated around 20,000 appm He. This study was evaluated the effect of He fluence and irradiation temperature on the bubble behaviors in Be12Ti and beryllium. The irradiation of He ion was carried out using an In-situ transmission electron microscope equipped with an ion gun by implantation up to 25,000 appm He. For comparison, beryllium was carried out in similar experiments.Bubbles were observed in Be12Ti at 873 K after the implantation of 15,000 appm He, while in Be at the same temperature, their appearance occurred post-implantation of 4,500 appm He. In addition, growth of bubbles in beryllium was observed with increase of He ion fluence. Sizes and densities of bubbles in Be12Ti and beryllium were evaluated irradiated in 25,000 appm He at 673 and 873 K.In the case of Be12Ti, fine bubbles were observed in a high density state, and it was found that the swelling rate was a few percent. By contrast, bubbles in beryllium had a diameter in the range from tens to several tens nanometers, and growth of these bubbles were observed. The swelling rate of beryllium was approximately 18 %.

Lanthanum plumbide as a new neutron multiplier material

A new neutron multiplier based on lanthanum plumbide LaPb3 was produced using two different casting methods. The argon-arc melted material consists of LaPb3 dendrites between grains of LaPb2 phase of approximately equal volume fraction. The induction melted material is composed primarily of large LaPb3 grains, about 80 μm in size. Corrosion testing at 20 and 300 °C in air revealed the rapid degradation of LaPb3 without the formation of a protective passivation layer. The corrosion process results in significant volumetric expansion accompanied by cracking. When LaPb3 is exposed to air at 500 °C, pure lead as a corrosion product melts to form layers protecting against rapid corrosion. LaPb3 showed low (113–193 MPa), but sufficient strength for functional use in the temperature range of 20–500 °C.

Tritium Breeding Ratio Evaluation of Solid Breeder Concepts for the FESS-FNSF

This paper presents a parametric study of the Fusion Energy System Studies-Fusion Nuclear Science Facility’s (FNSF’s) tritium breeding performance for several solid breeder concepts, neutron multiplying materials, and blanket materials, assuming volume fractions based on the most recent FNSF design as a realistically representative fusion facility. In this study, we initially surveyed the tritium breeding ratio (TBR) of several solid breeder concepts by employing a simplified but efficient one-dimensional (1-D) infinite cylinder reduced-order model (ROM). Parametric studies were performed with the ROMs for the full range of breeder-to-multiplier ratios to identify the optimum mixture compositions for each breeder type that would lead to a maximum TBR. These optimized breeder-multiplier combinations were then homogenized with FNSF blanket component materials to estimate their impacts on the TBR. Subsequently, as a validation step for the optimal designs, TBR calculations were performed using a more realistic modified 1-D ROM with inner and outer breeding regions, as well as with a fully detailed 22.5-deg three-dimensional (3-D) sector of the FNSF to assess the impact of geometry details on the TBR. The differences between the two 1-D models were negligible, while the ROMs were able to correctly predict trends and identify the maximum and minimum TBR cases, as well as show consistent biases relative to the results produced by the full 3-D, 22.5-deg sector for specific breeder/multiplier combinations. Solid breeder concepts such as , , and outperformed all others in this study in terms of TBR performance when combined with all the neutron multiplier materials selected. An underlying goal of this study was to develop and improve rapid and reliable ROMs to aid designers during parametric optimizations of highly complex and computationally expensive fusion models.

Initial Neutronics Investigation of a Chlorine Salt–Based Breeder Blanket

Tritium breeding blankets within D-T-fueled fusion reactors contain lithium compounds and typically require neutron multiplier materials to achieve a tritium breeding ratio (TBR) consistent with self-sustaining operation. Liquid breeder blankets have some advantages over solid blankets, and previous blanket studies have investigated liquid metal as well as liquid salt–based blankets. Liquid salts have reduced magnetohydrodynamic effects as compared to liquid metals, but typically have a lower TBR. Recently, advanced fission reactor concepts have considered chloride-based salts in their design, and there is a significant amount of research work occurring to study these salts. Chloride salts have previously been considered for fusion reactors, but studies have typically found lower breeding ratios than for fluoride salts, such as 2(LiF)-BeF2 (flibe) so they have not been further developed. In this work, we use a one-dimensional cylindrical radiation transport model of a conceptual tokamak fusion reactor to investigate the neutronics feasibility of using a chloride salt–based blanket that uses chlorine enriched in 37Cl, which has both a low capture cross section and a substantial (n,2n) cross section. It is found that chloride salts (LiCl mixed with BeCl2 and/or PbCl2) can potentially achieve a ~3% to 5% higher TBR than fluoride molten salts, notably flibe, in the absence of a solid multiplier. Including a solid multiplier, however, does narrow this advantage, with TBRs estimated within ~1% of flibe with a 2-cm Be multiplier. Chloride salts can also reach lower melting points than flibe, potentially improving the scope for the use of reduced activation ferritic-martensitic steel as a structural material. There is substantial uncertainty in the calculations driven by limited thermochemical data for the Cl salts, plus cross-section uncertainties. The production of 36Cl through 35Cl(n,g) and 37Cl(n,2n) has the potential to challenge the waste disposal rating of the blanket. Calculations indicate that, while this is not an immediate showstopper, this case depends upon the exact waste disposal rating criteria used for 36Cl. Further work could reduce these uncertainties with improved thermochemical data, higher-fidelity modeling for downselected salts, and more refined waste disposal calculations and regulatory guidance. Finally, it must be recognized that, as for all molten salts, corrosion and chemistry can present appreciable technical challenges that require further assessment in developing a practical blanket concept, and also that the enrichment of chlorine presents an additional technical and supply chain challenge.

Comprehensive investigation of lattice dynamics of neutron multiplier beryllides: vibrational spectra and thermal performances

In order to realize tritium self-sufficiency during the operation of fusion reactors, neutron multiplier materials are needed to multiply the number of neutrons. Beryllium (Be) and beryllides (Be12V, Be12Ti) are considered as candidates for neutron multiplier materials in the tritium breeding blanket. Thermal performances are essential data to evaluate the service performance of materials in future fusion reactors. Due to the toxicity of beryllium, the preparation and analysis of neutron multiplier materials are challenging. As a result, the performance data of neutron multiplier materials are limited. It is essential to adopt the theoretical method to get the basic parameters for the design of a tritium breeding blanket and optimization of neutron multiplier. In the present work, the electron structure, vibrational spectra, and thermal performances have been theoretically investigated to understand thermal expansion and thermal conductivity properties. Experimental data are also included to compare with the calculated results and detailed analysis is presented. Due to the reactivity with water, BeO will form on the surface of Be. Therefore, BeO was studied as well. The bulk modulus of Be is the smallest of the four materials, while BeO is the largest. Higher bulk modulus has a better ability to restrain the deformation due to pressure. The c-axis of the Be12V has the largest thermal expansion coefficient, while the c-axis of the Be12Ti is the smallest. It has been found that the larger the lattice constant of crystal axes, the smaller the linear thermal expansion coefficient in same material. Electrons play a major role in the heat transfer of Be12V and Be12Ti. At high temperatures, phonon and electron coupling must be considered. Isotopeenrichment and enlarging the grain size cannot significantly improve the phonon thermal conductivity when the temperature is higher than room temperature. The calculated values can be adopted to predict the materials’ performances and support future fusion reactors’ neutron multiplier material preparation.

Ultrasonic Anisotropy in Composites: Effects and Applications

Stiffness anisotropy is a natural consequence of a fibrous structure of composite materials. The effect of anisotropy can be two-fold: it is highly desirable in some cases to assure a proper material response, while it might be even harmful for the applications based on “isotropic” composite materials. To provide a controllable flexibility in material architecture by corresponding fibre alignment, the methodologies for the precise non-destructive evaluation of elastic anisotropy and the fibre orientation are required. The tasks of monitoring the anisotropy and assessing the fibre fields in composites are analyzed by using the two types of ultrasonic waves suitable for regular plate-shaped composite profiles. In the plate wave approach, the effect of “dispersion of anisotropy” has been shown to make the wave velocity anisotropy to be a function of frequency. As a result, the in-plane velocity pattern measured at a certain frequency is affected by the difference in the wave structure, which activates different elasticity against the background of intrinsic material anisotropy. Phase velocity anisotropy and its frequency dependence provide a frequency variation of the beam steering angle for plate waves (dispersion of beam steering). In strongly anisotropic composite materials, the beam steering effect is shown to provide a strong focusing of ultrasonic energy (phonon focusing). For bulk shear waves, the orthotropic composite anisotropy causes the effect of acoustic birefringence. The birefringent acoustic field provides information on stiffness anisotropy which can be caused by internal stresses, texture, molecular or/and fibre orientation. On this basis, a simple experimental technique is developed and applied for mapping of fibre orientation in composite materials. Various modes of acoustic birefringence are analyzed and applied to assessing the fibre fields in injection moulding composites and to identify the fibre lay-ups in multiply materials. The birefringence pattern is also shown to be sensitive and applicable to characterizing impact- and mechanical stress-induced damage in composites.

Neutronic analysis of effects of inboard materials on the size of a tokamak fusion reactor

• Inboard materials play a key role in determining the size of the reactor. • The reactor parameters were determined via coupled tokamak systems analysis. • Optimal combination of multiplier and reflector can lead to small reactor. The effects of inboard materials on the size of a tokamak fusion reactor were studied based on reactor system parameters determined through a coupled analysis of tokamak systems and neutron transport. With an inboard blanket, the shielding capability was the best for tungsten carbide (WC), showing a smaller reactor size than cases using other shield materials. Shield materials such as WC, borated steel (BS), and ferritic/martensite steel (FMS) showed improved neutron economy for tritium breeding in the blanket, which was attributed to neutrons reflected by tungsten (W) and steel (Fe). In a configuration with only an outboard breeding blanket, it was found that the tritium self-sufficiency requirement could be satisfied with an inboard blanket replaced by a neutron multiplier and/or reflector materials. Titanium carbide (TiC) showed the best neutron reflecting capability. With lead (Pb) as a neutron multiplier, neutron spectrum softening by a Pb multiplier enhanced the reflector performance and increased tritium breeding at the outboard blanket. Thus, an optimum combination of multiplier and reflector materials can lead to smaller reactor sizes than reactors with an inboard blanket.

English

Sweet potato (Ipomoea batatas (L.) Lam.) is a staple food that contributes to food security for communities in sub-Saharan Africa. The production of sweet potato in coastal Kenya is low and this has been associated with shortage of planting material during the rainy season. A study was therefore conducted at the Pwani University farm to assess the production of sweet potato planting material using cuttings planted under different methods. The methods that were evaluated in this study were: Planting in pits without lining, planting in pits with lining, planting on ground surface, planting in sacks without lining and planting in sacks with lining. A randomized complete block design was used, with factorial arrangement of treatments. Fifteen treatments were replicated in three blocks. In the long rains season, sweet potato in the treatments without lining produced longer vines than those with lining. Different vine planting methods produced similar vine lengths in the off-season multiplication. The planting methods without lining are recommended for use by farmers during the long rains season multiplication of planting material. It is also recommended that farmers start the multiplication during off-season and continue during the long rains season so as to multiply enough planting material that will allow expansion of the area under sweet potato. Key words: Food security, Ipomoea batatas, low production of sweet potato, off-season multiplication, vine planting methods.

Ab initio study of hydrogen diffusion in Be and Be12Ti for fusion applications

Beryllium and Be12Ti are considered as neutron multiplier materials in various tritium breeding blanket designs for future fusion reactors. Tritium and helium are formed due to neutron-induced nuclear transmutation and accumulated in beryllium-based materials. As a β-radioactive isotope, tritium represents additional safety concern with respect to sudden release during operation or hindering consequent recycling. Therefore, detailed study of its accumulation and release is indispensable. In particular, it is important to reveal which process controls tritium release: binding of tritium with traps or diffusion out of material bulk. Ab initio calculations of solution energies at all non-equivalent hydrogen interstitial sites confirm easier dissolution of tritium in Be12Ti in comparison with pure beryllium. The 17 considered diffusion barriers between adjacent hydrogen interstitial sites are calculated using a dimer method instead of a computationally more expensive nudged elastic band approach. Three possible diffusion paths in Be12Ti were revealed: two of them along the c-axis and another one along the a- and b-axes. Considering calculated hydrogen diffusion barriers in Be12Ti, one can conclude that at low hydrogen concentrations and low temperatures, diffusion occurs preferentially along the c-axis. The rate limiting diffusion barriers in Be12Ti are found to be in the range of 0.4–0.5 eV, which are comparable to 0.42 eV in pure beryllium suggesting that tritium diffusion in both materials proceeds at comparable rates. Therefore, binding of tritium atoms with vacancies and other lattice imperfections are expected to play a crucial role in tritium retention.

Assessment of multiscale hydrogen desorption models from (0001) Be surfaces

Beryllium is a proposed neutron multiplier material in future fusion devices like ITER and DEMO. As such, numerous studies of neutron-irradiated beryllium specimen have shown the appearance of facetted bubbles with considerable tritium accumulation at their (0001) surfaces. In this work, we thus simulate hydrogen isotope desorption from (0001) beryllium surfaces using a multiscale modeling approach. We first establish that very fast surface diffusion of hydrogen adsorbed in FCC and HCP sites can be studied independently from the bulk by means of static DFT calculations, as diffusion into the bulk is inhibited by multiple diffusion processes with an accumulated energy barrier of almost 3 eV. Static saddle point searching dimer DFT calculations also reveal energy barriers for associative desorption in a range between 0.94 eV and 1.52 eV due to varying amounts of co-adsorbed hydrogen isotopes in the immediate vicinity of the desorbing pair. A simple rate equation-based desorption model incorporating these associative desorption processes is constructed and compared to thermal programmed desorption (TPD) spectra of earlier exposure experiments. It is found that the simple rate equation-based desorption model seemingly already explains some characteristic aspects of the TPD spectra like a peak shift to lower temperatures by almost 100 K and considerable broadening with increasing coverage. To asses the merit of this model, a refined model effectively taking the repulsion among adsorbed hydrogen isotopes into account by means of relevant correlations from kinetic Monte Carlo (KMC) simulations using a cluster expansion (CE) truncation is considered as well. It is found that the simple model’s reproduction of characteristic features of experimental spectra is the result of error cancellation from considering an incomplete set of desorption processes on one hand and neglecting repulsion among adsorbed hydrogen on the other hand. We conclude that the experimental TPD spectra can not be fully explained by either of the two models and further research particularly considering additional adatoms like oxygen are necessary to understand the TPD spectra of the corresponding experiments.

Development and characterization of advanced neutron multiplier materials

Neutron multiplier materials are essential for self-sufficient tritium production and are closing the fuel cycle of fusion reactors. Until now, a concept of pebble bed consisting of interchanging layers of beryllium and lithium ceramic pebbles was considered for the Helium-Cooled Pebble Bed (HCPB) tritium-breeding module of the first experimental fusion reactor ITER as well as for the next demonstration fusion reactor DEMO. However, this concept depends on the availability of large amounts of pure beryllium pebbles and is also limited by its material properties like for example the tritium accumulation under irradiation.The results of tritium retention and analytical microstructural studies of beryllium pebbles obtained within the framework of the HIDOBE irradiation campaign suggest that a significant fraction of generated tritium (up to 100% below 500°C) is trapped within helium bubbles. Being negligible in the ITER tritium-breeding module (TBM), the total accumulated tritium inventory imposes severe safety issues and exceeds acceptable limits for the DEMO blanket. Therefore, advanced neutron multiplier materials such as beryllides have to be well characterized for their applicability in the HCPB blanket of DEMO and beyond. The usage of an advanced material with lower volumetric swelling, lower tritium retention, increased irradiation and chemical resistance as well as with higher melting temperature allows to switch from the pebble bed concept to a solid hexagonal block-based one. For the fabrication of titanium beryllide samples both the semi-industrial fabrication route utilizing the hot extrusion of rods and the industrial approach using vacuum hot pressing of a Be-Ti powder mixture were explored. In this contribution, we discuss reasons for the transition from pure beryllium to beryllides, respective changes of the HCPB blanket design and a successful demonstration of the feasibility of a beryllide block fabrication by an industrial method.

Tritium release and retention in beryllium and titanium beryllide after neutron irradiation up to damage doses of 23-38 dpa

• For the first time, beryllium and titanium beryllide as massive pellets were irradiated in a nuclear reactor at temperatures of 710−1040 K up to production of 430–653 appm tritium and 4144–5992 appm helium, respectively. • The tritium retention in irradiated Be-7at.%Ti pellets is much lower than that in Be pellets, for example, 1 and 0.5 % to 48 and 34 % at 940 and 1040 K, respectively. • Open channels along grain and sub-grain boundaries are formed under neutron irradiation in beryllium, essentially contributing to the enhanced tritium release. Titanium beryllide is considered as an advanced neutron multiplier in the Helium Cooled Pebble Bed (HCPB) blanket of DEMO. Neutron irradiation of titanium beryllide together with beryllium as a reference material in material testing nuclear reactors can give essential data for the DEMO blanket design. Be-7at.%Ti (Be-7Ti) as well as Be were irradiated in the HFR, Petten, the Netherlands, at four temperatures of 710, 800, 940, 1040 K up to 23, 31, 36, 38 dpa, respectively. The post-irradiation examination (PIE) included thermal-programmed desorption (TPD) and Vickers hardness tests as well as microstructure study by optical metallography. Be and Be-7Ti pellets maintained their integrity after irradiation. Microstructure of Be-7Ti contains two phases, mainly, TiBe 12 , and also small amount of Be. Under irradiation, gas bubbles were formed in Be samples as well as in Be-phase in Be-7Ti samples. These bubbles contain helium and tritium produced in Be under irradiation. TPD tests showed a much lower tritium retention in Be-7Ti than in Be for all four irradiation temperatures. Vickers hardness of TiBe 12 phase is much higher than that of Be-phase. According to the obtained data, Be-7Ti could be considered more preferred than Be as a neutron multiplier material in future fusion reactors due to the enhanced radiation damage resistance.

Oxidation behaviour of neutron irradiated Be pebbles

Abstract Beryllium will be used as neutron multiplier material to increase the tritium breeding ratio in fusion reactors. The chemical reactivity of beryllium is a factor of major concern since the oxidation of beryllium at temperatures above 800°C can become an uncontrollable process. The present work presents a detailed study of chemical composition and reactivity of Be pebbles after exposure to neutron irradiation, up to 3000 appm He production, during the HIDOBE-01 campaign and their influence in the oxidation kinetics process. In order to get information on this reaction, samples were oxidized at 700°C under controlled air atmosphere (60% humidity) and with a mixture of 60% O2 + 40% N2. The chemical composition of the irradiated samples was studied using ion beam analysis and scanning electron microscopy. The observations revealed the presence of small flakes which are distributed in the surface, however no modifications in the pebbles shape after longer oxidation were observed. Ion beam elemental maps confirmed the presence of Be oxides distributed on the surface. Moreover, the width of the oxide layer increases with the annealing time, following a parabolic law.

Design of Two Experimental Mock-Ups as Proof-of-Concept and Validation Test Rigs for the Enhanced EU DEMO HCPB Blanket

Within the framework of EUROfusion activities, the helium cooled pebble bed (HCPB) breeding blanket is under development in Karlsruhe Institute of Technology (KIT). The enhanced HCPB, using a fissionlike fuel breeder pin assembly configuration, is proposed as the near-term breeding blanket for the European Union DEMOnstration power plant (EU DEMO). The helium gas with a pressure of 8 MPa is used as the coolant, EUROFER is used as the structural material, advanced ceramic breeder pebbles are used as the tritium breeder, and Be12Ti is used as the neutron multiplier material. In contrast to the former HCPB cooling plate configuration, the fuel breeder pin assemblies greatly reduce the pressure drop, reducing the circulating power to the level where state-of-the-art helium turbomachinery can be used. However, because of the reduced coolant velocity in the breeder zone, the heat transfer performance is compromised, especially in the annular channel of the fuel breeder pins. An increased surface roughness is therefore proposed as a heat transfer augmentation technique for the fuel breeder pins. Although heat transfer augmentation using artificial roughness is common, it is relatively novel for small annular gaps with moderate velocity as the ones in the fuel breeder pins. Currently, a dedicated correlation for the small annular rough-wall channel is available to predict the Nu number. This correlation is wished here to be benchmarked and validated experimentally. Therefore, an experimental investigation on a fuel breeder pin mock-up (mock-up 1) is planned. Additionally, based on computational fluid dynamics calculation, unsteady, nonuniform flow patterns were found at the return flow after the jet impingement in the first-wall region. Another upscaled mock-up (mock-up 2) to investigate the nonuniform flow patterns of the return flow in the annular channel of the fuel breeder pin is planned. The dedicated experimental campaigns are foreseen at the Helium Loop Karlsruhe (HELOKA) in KIT as validation and proof-of-concept test rigs for this enhanced pin design. In this paper, the motivation and the preliminary design of these two mock-ups of the enhanced EU DEMO HCPB blanket are shown, together with the plan for the foreseen experiments.

Learning curve with input price for tracking technical change in the energy transition process

Abstract Learning curve has been recognized as an effective tool for tracking the technological change and widely used in assisting policy design for energy technologies. But uncertainties underlying input price and learning rate would greatly influence the accuracy and reliability of learning curves. Regression to a learning curve using lead-acid price and production (between 1989 and 2013) yields a poor fit performance. This study then integrates cost decomposition with learning curve to remove the influences from input price changes in lead-acid batteries. Total lead-acid batteries' cost is decomposed into materials cost and residual cost. Materials cost is calculated through multiplying material consumption and its price to isolate the influences from input price changes. Learning curve is then constructed to trace the declines of materials consumption and residual cost. Regression with the improved learning curves performs better than a standard one, with the fit performance improved greatly. The study also indicates a learning effect in the material consumption in the long term. Integrating cost decomposition into learning curve is then proved to be a meaningful trial for mitigating the bias from input price changes. Learning rates of mature technologies such like lead-acid batteries indicate the importance to understand the cost reduction in them, which might significantly affect the policies’ performances for emerging ones.

First principles review of options for tritium breeder and neutron multiplier materials for breeding blankets in fusion reactors

The current breeding blankets proposed in the different conceptual fusion power plants are based mainly on the use of Li4SiO4 and/or Li2TiO3 as tritium breeder and Be/Be12Ti as neutron multiplier or an eutectic Li17Pb83 for as a hybrid tritium and neutron multiplier. While these materials offer some tritium breeding capabilities, some recent studies show that the tritium self-sufficiency may not be ensured with these materials due to the strong reduction of blanket coverage after the integration of other in-vessel reactor systems (heating and current drive, limiters, large or double-null divertor systems, etc.). Also, some materials like Be raises several key feasibility concerns. The goal of this paper is to perform an update of the screening for tritium breeder and neutron multiplier materials and to assess the tritium breeding performance of the selected compounds in order to reveal new options. As for the neutron multiplier materials, a new subdivision between solid and liquid multipliers is proposed. For the selected compounds, detailed 3D heterogeneous neutronic analyses have been performed with MCNP5-1.60 assuming the architecture of the current EU DEMO Helium Cooled Pebble Bed (HCPB) as a benchmark breeding blanket. From the point of view of ceramic breeders, Li8ZrO6 has been found to outperform Li4SiO4 by more than 4% in terms of tritium breeding, having 6% higher melting point. From the point of view of solid neutron multipliers, Be12Cr, Be12V, Be13Zr and Be13Y show a similar performance as Be12Ti, while LaPb3, Zr5Pb4 and YPb2 offer a solution for a Be-free blanket. As for liquid multipliers, Pb in combination with a ceramic breeder shows a very promising option. Moreover, Pb compounds like Pb90Mn10 and Pb95Ba5 offer similar performance as Pb with a lower melting point (290 °C). Due to the significant advantages of molten Pb as neutron multiplier, future work will be conducted to define a design of a helium cooled Molten Lead Ceramic Breeder blanket, as simple, cost effective blanket concept.

A biolistic-based genetic transformation system applicable to a broad-range of sugarcane and energycane varieties

ABSTRACTSugarcane and energycane (Saccharum spp. hybrids) are prominent sources of sugar, ethanol, as well as high-value bioproducts globally. Genetic analysis for trait improvement of sugarcane is greatly hindered by its complex genome, limited germplasm resources, long breeding cycle, as well as recalcitrance to genetic transformation. Here, we present a biolistic-based transformation and bioreactor-based micro-propagation system that has been utilized successfully to transform twelve elite cane genotypes, yielding transformation efficiencies of up to 39%. The system relies on the generation of embryogenic callus from sugarcane and energycane apical shoot tissue, followed by DNA bombardment of embryogenic leaf roll discs (approximately one week) or calli (approximately 4 weeks). We present optimal criteria and practices for selection and regeneration of independent transgenic lines, molecular characterization, as well as a bioreactor-based micro-propagation technique, which can aid in rapid multiplication and analysis of transgenic lines. The cane transformation and micro-propagation system described here, although built on our previous protocols, has significantly accelerated the process of producing and multiplying transgenic material, and it is applicable to other varieties. The system is highly reproducible and has been successfully used to engineer multiple commercial sugarcane and energycane varieties. It will benefit worldwide researchers interested in genomics and genetics of sugarcane photosynthesis, cell wall, and bioenergy related traits.

Ab initio study of Be and Be12Ti for fusion applications

The paper applies ab initio methods for investigation of possible origin of the easier tritium release from titanium beryllide Be12Ti as compared with that of beryllium. Both materials are considered as neutron multiplier materials in some design concepts of tritium breeding blanket of future fusion reactor. The only stable tetragonal crystal structure of Be12Ti was studied in this work. Hydrogen solution energy in all non-equivalent interstitial positions of Be12Ti was found to be lower than that for beryllium suggesting easier dissolution of hydrogen atoms in Be12Ti. Maximal binding energy of hydrogen atoms in vacancy linearly decreases with the number of hydrogen atoms both in beryllium and Be12Ti. The highest hydrogen binding in a vacancy is 0.4 eV lower in Be12Ti, predicting easier release of hydrogen from beryllide with respect to beryllium. No formation of chemical bonds between hydrogen atoms inside both beryllium and Be12Ti vacancies was observed.