Fusion Reactor Blanket Research Articles

Preparation of High Sphericity Li<sub>2</sub>TiO<sub>3</sub> Tritium Breeder by Polymer Assisted Sedimentation Method

The tritium breeder materials used in solid tritium breeding blanket of fusion reactor are lithium-based ceramics, like lithium titanate (Li2TiO3), lithium orthosilicate (Li4SiO4), lithium aluminate（LiA1O2) and lithium zirconate (Li8ZrO6). Among them, Li2TiO3 has the advantages of stable chemical property, high mechanical strength and high lithium content, so it has become one of the preferred materials. Based on the design of solid blanket and tritium extraction process, breeder materials need to be made into a certain shape of pebbles. B Spherical breeder is widely used in the design of fusion reactor for the larger specific surface area, more channels between the pebbles to diffuse and release tritium, and easier loading or unloading. However, due to the high requirements of the breeder pebbles, the production control are not complete. In this paper, a new preparation method of ceramic pellets with high sphericity, polymer assisted sedimentation method, was used to prepare ceramic pebbles. The slurry consists of micro-scaled ceramic powder and organic monomers. Spherical droplets of ceramic slurry were obtained by sedimentation method in viscous liquids. The organic monomer contained in the droplets was heated to polymerize into small pellet preforms during the sedimentation process. By optimizing the sintering process, the fabricated pebbles could perform high compressive load (>18 N), spherical shape (0.95~1.02), uniform particle size (0.8~1.2mm) and porosity (15~20 %). The main advantages of this method is the mild preparation conditions (50~110 °C), the simplified transfer and purification process and convenient industrialization.. The obtained product has high spherical shape, uniform particle size, high compactness and mechanical strength.

Peculiarities of Heat Transfer in Flow of a Liquid Metal in Vertical Duct in Coplanar Magnetic Field

The hydrodynamics and heat transfer at downward and upward flows of liquid metal in the rectangular duct with the height-to-width aspect ratio of ~3/1 in a coplanar magnetic field were studied. The problem simulates the flow in head-transfer ducts of the liquid metal test module of the blanket of a fusion reactor of tokamak type. The experiments were conducted on the basis of the mercury magnetohydrodynamic test bench. The probe technique was used for measurements in the flow. The averaged profiles of velocity, temperatures, temperature oscillations of flow, and duct wall temperatures were measured in the duct cross section located far from the beginning of heating in the region of uniform magnetic field. The magnetic field suppresses the turbulent transfer owing to the decrease in the heat transfer coefficients and the increase in the wall temperature. However, under the conditions of downward flow, a significant influence of the counter thermogravitational convection (TGC) on the flow was discovered. The interaction of TGC with an external magnetic field in a number of modes leads to the appearance and development of secondary flows. The largescale vortex structures of TGC generate low-frequency temperature oscillations of very high intensity. In this case, the heat emission is improved.

Joint inhomogeneity in dissimilar friction stir welded martensitic and nanostructured ferritic steels

ABSTRACTDissimilar welding between oxide dispersion strengthened ferritic (ODS) steel and reduced activation martensitic steel would be required for constructing the advanced blanket of progressive fusion reactors. In this study, we achieved dissimilar joints by friction stir welding, and aimed to characterise and ameliorate joint inhomogeneity. Main results reveal that the joint inhomogeneity is generated from discrepant microstructural evolutions within the martensitic and ODS ferritic steels. The ODS steel achieves evolution by the dynamic recrystallisation, while the martensitic steel undergoes phase transformation that drastically hardens the stir zone. By a proper post-weld heat treatment, the joint inhomogeneity can be effectively ameliorated due to carbide reprecipitation and stress relief in the joint.

Convection instability in a downward flow in a vertical duct with strong transverse magnetic field

The downward flow in a vertical duct with one heated and three thermally insulated walls is analyzed numerically using the two-dimensional approximation valid in the asymptotic limit of an imposed strong transverse magnetic field. The work is motivated by the design of liquid metal blankets with poloidal ducts for future nuclear fusion reactors, in which the main component of the very strong magnetic field is perpendicular to the flow direction and very strong heating is applied at the wall facing the reaction chamber. The flow is found to be steady-state or oscillating depending on the strengths of the heating and magnetic field. A parametric study of the instability leading to the oscillations is performed. It is found among other results that the flow is unstable and develops high-amplitude temperature oscillations at the conditions typical for a fusion reactor blanket.

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.

Efficient hydrogen and deuterium permeation reduction in Al2O3 coatings with enhanced radiation tolerance and corrosion resistance

One of the major bottlenecks in the development of the breeding blanket of the DEMO fusion reactor is the suppression of tritium permeation through structural steels as well as their protection against dissolution-corrosion by interaction with high temperature heavy liquid metals. Reduction of tritium permeation and corrosion of structural steels are crucial issues in order to enhance reactor safety and avoid operational implications. As a solution to these two daunting challenges, we developed multifunctional alumina coatings capable to tackle, at the same time, tritium permeation and Pb–Li corrosion. The coatings are deposited by pulsed laser deposition and are essentially amorphous, with nanocrystalline inclusions. By optimizing the deposition process, we provide experimental evidence that PLD-grown alumina yields a permeation reduction factor for hydrogen and deuterium well above 104, even after electron irradiation, and a suitable protection against corrosion by the Pb–16Li eutectic. Given these results, the multifunctional, PLD-grown alumina coatings stand out as a viable solution to some of the long-lasting issues related to fusion technologies.

Effect of isotope enrichment on performance of lead-lithium blanket of inertial fusion reactor

Simulation studies on neutron transport and material activation for lead-lithium (Pb-17Li) blanket of the inertial fusion reactor “KOYO-FAST” were performed with the PHITS code and the EASY 2005 code. Tritium breeding ratio (TBR) and productions of harmful radioactive isotopes 210Po and 203Hg in the blanket are made clear. The blanket performances improved by enrichment of stable isotope (6Li, 204Pb, 206Pb or 207Pb) are also made clear. The TBR is 1.28 for the blanket with a natural Pb-17Li alloy. The large TBR is allowed because of large blanket space in the reactor chamber. The TBR reaches 1.64 when 6Li is enriched to 90%. In this case, intermediate and thermal neutrons, that are necessary to produce 210Po and 203Hg, are preferentially captured by the 6Li (n, α) T reaction. Then, the productions of 210Po and 203Hg are mitigated. The enrichment of 204Pb leads to a small TBR, because neutrons are captured by 204Pb due to its large cross section for (n. γ) reaction. The TBR of 206Pb or 207Pb enriched Pb-17Li is more or less the same with that of the natural Pb-17Li. In these cases, the productions of 210Po and 203Hg are slightly mitigated because of the removal of 208Pb and 204Pb from the blanket.

Experimental Study on the Natural Circulation Subcooled Boiling of Square Channel in Fusion Reactor Blanket

Blanket is implemented in fusion reactor for removing heat, shielding neutrons, and tritium breeding. Under accident conditions, natural circulation flow will emerge due to the buoyancy forces caused by density gradient in the different parts of the loop. Without the function of moving machinery, the natural circulation will occur to remove the heat of the nuclear reactor core, which denotes natural circulation is one of the significant passive cooling mechanisms of a new reactor. The appearance of subcooled boiling will greatly affect density gradients in the closed loop and lead to the flow instability of coolant. Experimental study has been performed to investigate natural circulation subcooled boiling characteristics in square channels of fusion blanket. Experimental results show that the natural circulation mass flow ascends with the increase of the heat flux and the decline of pressure, and subcooled boiling is easier to appear in natural circulation because natural circulation flow is usually small and heat transfer intensity is lower compared to forced circulation. The location of onset of nucleate boiling (ONB) moves to inlet of test section and the wall superheat increases with the ascend of heat flux. A new empirical experimental correlation is developed on the foundation of experimental data, which correlates heat flux at ONB in natural circulation with the maximum relative error of 35%. Furthermore, the correlations of wall superheat, location of ONB, and subcooled boiling heat transfer coefficient are also acquired to predict the occurrences of natural circulation subcooled boiling of square channel.

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.

Strain evaluation using a non-contact deformation measurement system in tensile tests of irradiated F82H and 9cr ODS steels

We developed a non-contact deformation measurement system to accurately evaluate strain for post irradiation tensile testing, since conventional strain gages cannot be used for small size specimens. The strain calculated from cross-head displacement generally includes deformation from specimen shoulders, fixtures, and the test frame in addition to the deformation from the specimen gauge section. In our system, the distance between painted marks within the specimen gauge section was measured using a high resolution video camera to evaluate the specimen deformation during room temperature tensile testing. The test materials were F82H and 9Cr ODS steels irradiated up to ≈71 displacements per atom (dpa) at about 573 K in High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). Our system yielded accurate stress strain curves without deformations other than the specimen gage section, and the elongation was less than that calculated from cross-head displacement. This system can contribute to expanding the technically reliable database for the design activity of fusion reactor blanket, including the effects of irradiation on tensile properties.

Application of various observational techniques to the characterization of MX in F82H

ABSTRACT Reduced activation ferritic/martensitic steels (RAFMs) have been developed for the structural material of fusion reactor blanket. Extensive researches were conducted using transmission electron microscopy (TEM) to define the precipitation behavior of MX (M: Metallic elements, X: C and/or N) precipitates such as vanadium and tantalum carbides and/or nitrides, since they strongly affect material properties. However, it is not completely defined yet. TEM can obtain detailed information, but the observable volume is limited and it needs a long time to obtain representative and statistically reliable microstructural information. It is fruitful to complement TEM results applying other characterization techniques together. In this work, we studied the precipitation behavior of MX in the RAFM, F82H, through the application of various techniques with larger observable volumes and compared with TEM past results. We additionally tried to find sub-nanometric MX precipitates that can be hardly observed through TEM. From the result, the TaX (X: C and/or N) phase surely varied depending on the chemical composition of alloy and heat treatment; some TaX precipitates were unstable during tempering, MX precipitates were intergranular, and sub-nanometric MX were not found in the matrix. These obtained results were helpful to define the precipitation behavior of MX in F82H.

Microstructure and phase evolution of Li4TiO4 ceramics pebbles prepared from a nanostructured precursor powder synthesized by hydrothermal method

Li4TiO4 ceramics pebbles have been proposed to be a promising tritium breeder materials for the fusion reactor blanket of the International Thermonuclear Experimental Reactor (ITER). In the present study, the hydrothermal method was first used to synthesize the nanostructured precursor powders for fabricating the Li4TiO4 ceramic pebbles. The precursor powders were composed of LiOH, Li2CO3 and Li2TiO3 with a grain size of 30–80 nm. Moreover, the microstructure and phase evolution of Li4TiO4 ceramics pebbles were investigated. The results indicated that the formation temperature of Li4TiO4 by hydrothermal method was lower than that of other methods. The pure phase Li4TiO4 ceramic pebbles with small grain size (average value 0.50 μm), satisfactory crush load (average value 41 N) and reasonable pore structure was sintered at 750 °C for 4 h in vacuum. This study provides a new method for preparing a high quality Li4TiO4 ceramic pebble, which has potential application as a tritium breeding material.

Experimental Study on Natural Circulation Heat Transfer of Square Channel in Water-Cooled Blanket

Square channel is widely used in the conceptual design of water-cooled blanket of fusion reactor for cooling and providing appropriate inner temperature field for tritium breeding. Blanket is one of the most important safety components, and the thermal-hydraulic characteristics of blanket directly determine the heat transfer efficiency and safe operation of fusion reactor. Under accident conditions, the natural circulation phenomenon occurs without any mechanical devices intervention when the field forces acting on the fluid produce density gradients able to induce natural convection, which are the main heat transfer mechanisms and the important measures to mitigate the consequence of the reactor accident. For square channel in blanket, the experimental study of natural circulation heat transfer has been conducted. It can be found that natural circulation mass flow is not an independent parameter, which increases with the increase in the heat flux and with the decrease in system pressure within the experiment scope. A correlation that can correlate the experimental data with the maximum deviation of 13% is established to predict natural circulation mass flow. Simultaneously, natural circulation heat transfer is strongly affected by system pressure and heat flux. And new correlations are developed on the foundation of experimental data to predict the heat transfer coefficients of natural circulation under different conditions. Comparing the experimental results with the results of forced circulation, it can be found that the heat transfer intensity of natural circulation is lower than that of forced circulation.

Effect of preheating on welding cold crack sensitivity of China low activation martensitic steel

The China low activation martensitic (CLAM) steel has been chosen as the primary structure material in Chinese series PbLi blankets for fusion reactors. The assembly preparation of the blanket would be seriously influenced by the weld performance of the CLAM steel. In this study, the welding cold crack sensitivity index of the CLAM steel was analyzed, then the maximum hardness in weld heat-affected zone (HAZ) and y-groove tests were performed. Based on the analysis to the maximum hardness in weld HAZ and the cracking rate of y-groove test at different preheating conditions, the results indicated that the HAZ maximum hardness of CLAM increased at firstly and then decreased with the increase of the preheating temperature (PT). The cracking rate of y-groove test reduced with the increase of the PT, and the crack rate maintained at a relative stable stage when the PT exceeded 250 °C. It revealed high welding crack tendency of the CLAM steel. Thus, the preheating before welding should be applied, and a PT at about 250 °C was proposed in order to prevent welding cracking.

Magnetohydrodynamic pressure drop and flow balancing of liquid metal flow in a prototypic fusion blanket manifold

Understanding magnetohydrodynamic (MHD) phenomena associated with the flow of electrically conducting fluids in complex geometry ducts subject to a strong magnetic field is required to effectively design liquid metal (LM) blankets for fusion reactors. Particularly, accurately predicting the 3D MHD pressure drop and flow distribution is important. To investigate these topics, we simulate a LM MHD flow through an electrically non-conducting prototypic manifold for a wide range of flow and geometry parameters using a 3D MHD solver, HyPerComp incompressible MHD solver for arbitrary geometry. The reference manifold geometry consists of a rectangular feeding duct which suddenly expands such that the duct thickness in the magnetic field direction abruptly increases by a factor rexp. Downstream of the sudden expansion, the LM is distributed into several parallel channels. As a first step in qualifying the flow, a magnitude of the curl of the induced Lorentz force was used to distinguish between inviscid, irrotational core flows and boundary and internal shear layers where inertia and/or viscous forces are important. Scaling laws have been obtained which characterize the 3D MHD pressure drop and flow balancing as a function of the flow parameters and the manifold geometry. Associated Hartmann and Reynolds numbers in the computations were ∼103 and ∼101-103, respectively, while rexp was varied from 4 to 12. An accurate model for the pressure drop was developed for the first time for inertial-electromagnetic and viscous-electromagnetic regimes based on 96 computed cases. Analysis shows that flow balance can be improved by lengthening the distance between the manifold inlet and the entrances of the parallel channels by utilizing the effect of flow transitioning to a quasi-two-dimensional state in the expansion region of the manifold.

Influence of friction stir welding conditions on joinability of V-alloy/SUS316L dissimilar joint

As one of the joining method to join between V-alloy and austenitic stainless steel SUS316L without melting of V and Fe to realize the self-cooled liquid-lithium blanket of fusion reactor, friction stir welding (FSW) is selected and the effect of FSW conditions on joinability of this dissimilar joint was examined. The sound dissimilar joint can be fabricated under the condition that V-alloy is set on the advancing side and the FSW tool is plunged into the SUS316L. Although the surface appearances of joint indicate that the case of the largest total heat input is the best condition, the results of cross-sectional microstructures and tensile tests suggests that the amount of adhesion to FSW tool seems to increase due to the higher temperature during FSW and the stir condition would be deteriorated. In addition, it is found that the higher rotational and traveling speeds might cause an insufficiency of the plastic flow and then crack or kissing bond would be generated. Moreover, it is revealed that one of the most preferable methods to prevent the adhesion of the mixture of base materials to the tool during processing is the selection of an appropriate coating on FSW tool.

Numerical simulation and experimental research on temperature and stress fields in TIG welding for plate of RAFM steel

The Reduced Activation Martensitic/Ferritic steel (RAFM steel) is known as one of the most important materials for future fusion reactor blanket and the welding process is unavoidable for blanket structure, so it is extremely necessary to research his welding performance. Some simulations and experiments about welding processes of RAFM steel have been done in this paper. The multi-welding processes for medium thickness plate of RAFM steel were simulated in three-dimensional finite element models by ANSYS software. The temperature fields and stress fields from simulation were contrasted with that from experiments respectively, and the heat affected zones of welding were calculated and checked. Meanwhile, the metallic texture graphic analysis for welded zones was performed. The results from both are very similar, and the fused metal welding zones are very well without obvious defects by used 316L welding padding. The hardness on welding zone is much higher than the heat-affected zone, also base materials. All of them can provide beneficial reference for blanket design and processing of Chinese Fusion Test Reactor.

Ratcheting behavior of Eurofer97 at 550 °C

The components in reactor are usually exposed to cyclic loading caused by frequent startups, shutdown and load fluctuations which lead to ratcheting effect. Eurofer97 is a structural material candidate for in-vessel components of future fusion reactor. To investigate the ratcheting behavior of Eurofer97, uniaxial stress-controlled cyclic loading tests are performed at 550 °C. Ratcheting rates under various loading conditions, including various peak stresses, stress ratios and stress rates are determined to build the database of Eurofer97 for future application in the blankets of fusion reactor. An already developed unified visco-plastic constitutive model for mod.9Cr-1Mo FM steels is further applied to adapt the ratcheting behavior of Eurofer97. The developed model describes both cyclic softening in strain-controlled LCF tests and its impact on the ratcheting behavior of mod.9Cr-1Mo FM steel fairly good, at both room temperature and 550 °C. Verifying whether the current model can also describe the ratcheting behavior of Eurofer97 is a further aim of the work reported here.

The effect of wall flow velocity on compatibility of high-purity SiC materials with liquid Pb-Li alloy by rotating disc testing for 3000 h up to 900 °C

The lead-lithium (Pb-Li) alloy is considered as a coolant and a tritium breeder for fusion reactor blanket systems. One of the critical requirements in realization of the systems is the compatibility of silicon carbide (SiC) and its composites, as structural and/or functional materials of the blanket, with liquid Pb-Li flow at high temperatures. This study aims to clarify the corrosion process of high-purity SiC materials under flow condition of liquid Pb-Li for 3000 h at 700 and 900 °C, respectively, by using the rotating disc corrosion equipment. Furthermore, the effect of wall flow velocity on the compatibility was specifically evaluated. For that purpose the decrease rate (about 32–37%) of the wall flow velocity on disc type’s sample by the dragging effect by fluidic viscosity was first calculated by numerical analysis of the rotating flow condition. The wall flow velocity dependence on the reaction layer’s thickness was then evaluated for both high purity SiC and SiC/SiC composites. The test results show that the formation process of the surface reaction layer mainly depends on both the time required for the chemical reaction and the wall flow velocity in the initial stage but the reaction layer thickness tended to approach to the constant regardless of the wall flow velocity with increasing exposed time. In short, no accelerated corrosion was found for about 3000 h. It was also found that there was no significant difference of the temperatures between 700 and 900 °C. Finally the key corrosion mechanism, i.e., the reaction of oxide impurities, e.g., Li2O, in liquid Pb-Li with the Si-oxide layer over SiC, was identified.

A promising tritium breeding material: Nanostructured 2Li2TiO3-Li4SiO4 biphasic ceramic pebbles

As an advanced tritium breeder material for the fusion reactor blanket of the International Thermonuclear Experimental Reactor (ITER), Li2TiO3-Li4SiO4 biphasic ceramic has attracted widely attention due to its merits. In this paper, the uniform precursor powders were prepared by hydrothermal method, and nanostructured 2Li2TiO3-Li4SiO4 biphasic ceramic pebbles were fabricated by an indirect wet method at the first time. In addition, the composition dependence (x/y) of their microstructure characteristics and mechanical properties were investigated. The results indicated that the crush load of biphasic ceramic pebbles was better than that of single phase ceramic pebbles under identical conditions. The 2Li2TiO3-Li4SiO4 ceramic pebbles have good morphology, small grain size (90 nm), satisfactory crush load (37.8 N) and relative density (81.8 %T.D.), which could be a promising breeding material in the future fusion reactor.