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.     

Experimental study and analysis of the purge gas pressure drop across the pebble beds for the fusion HCPB blanket

The lithium ceramic and beryllium pebble beds of the breeder units (BU), in the fusion breeding blanket, are purged by helium to extract the bred tritium. Therefore, the objective of this study is to support the design of the BU purge gas system by studying the effect of pebbles diameter, packing factor, pebble bed length, and flow inlet pressure on the purge gas pressure drop. The pebble bed was formed by packing glass pebbles in a rectangular container (56 mm × 206 mm × 396 mm) and was integrated into a gas loop to be purged by helium at BU-relevant pressures (1.1–3.8 bar). To determine the pressure drop across the pebble bed, the static pressure was measured at four locations along the pebble bed as well as at the inlet and outlet locations. The results show: (i) the pressure drop significantly increases with decreasing the pebbles diameter and slightly increases with increasing the packing factor, (ii) for a constant inlet flow velocity, the pressure drop is directly proportional to the pebble bed length and inlet pressure, and (iii) predictions of Ergun's equation agree well with the experimental values of the pressure drop.     

Fabrication of a Li4SiO4–Pb tritium breeding material

PbO₂-doped Li₄SiO₄ pebbles were successfully fabricated by a liquid-atmosphere sintering process. Those pebbles sintered at 1000 °C under atmospheric conditions were found to have an average diameter of 1.05 mm, a sphericity of 98%, a theoretical density of 90.9%, an average crush load of 24.3 N, and a main phase structure of Li₄SiO₄ with a small percentage of Li₈PbO₆. Subsequent optimization of this fabrication process yielded ceramic pebbles suitable for tritium breeding in a test blanket module (TBM).     

Fabrication and characterization of Li4SiO4 pebbles by melt spraying method

The solid breeder blanket concept proposed by the China features the tritium breeding ceramic as pebble beds in several submodules. The lithium orthosilicate (Li₄SiO₄) is considered as first candidate ceramic breeder materials fabricated by the melt spraying method, which is favorable to other processes in terms of density and recycling. The production process involves rapid quenching of the liquid droplets from the melt to room temperature which cause internal stresses and leads in some cases to formation of microcracks and the dispersion of mechanical properties. Molar ratio (Li/Si) of the pebbles was evaluated by ICP–OES. It is shown that the Li/Si ratio of the pebbles is slightly varying from batch to batch because of evaporation of lithium at high temperatures. The crush tests on single pebbles show that a mean value of 7.0 N was obtained in crush load experiments of 40 pebbles with a diameter of 1.0 mm. It results that heat treatment of pebbles improves the density and mechanical stability. The activation characteristics for the current composition of Li₄SiO₄ pebbles were assessed. The calculations were used to identify critical amounts of impurities and were compared to the results of pure material without impurities.     

Fabrication of Li2TiO3 pebbles by a freeze drying process

Li₂TiO₃ pebbles were successfully fabricated by using a freeze drying process. The Li₂TiO₃ slurry was prepared using a commercial powder of particle size 0.5–1.5 μm and the pebble pre-form was prepared by dropping the slurry into liquid nitrogen through a syringe needle. The droplets were rapidly frozen, changing their morphology to spherical pebbles. The frozen pebbles were dried at ?10 °C in vacuum. To make crack-free pebbles, some glycerin was employed in the slurry, and long drying time and a low vacuum condition were applied in the freeze drying process. In the process, the solid content in the slurry influenced the spheroidicity of the pebble green body. The dried pebbles were sintered at 1200 °C in an air atmosphere. The sintered pebbles showed almost 40% shrinkage. The sintered pebbles revealed a porous microstructure with a uniform pore distribution and the sintered pebbles were crushed under an average load of 50 N in a compressive strength test. In the present study, a freeze drying process for fabrication of spherical Li₂TiO₃ pebbles is introduced. The processing parameters, such as solid content in the slurry and the conditions of freeze drying and sintering, are also examined.     

Tokamak active laser pyrometry for tungsten deposited layer characterisation

In modern fusion reactors, the erosion of plasma facing surface results in layers deposition on tokamak “cold” surfaces. To provide efficient operation of tokamaks, it is essential to characterise the deposited layer with high tritium content. In situ rapid surface characterisation without reactor components disassembly is required. Active laser pyrometry together with a repetition rate Nd–YAG laser (1 Hz–1 kHz repetition rate frequency) applied for surface heating can be used to characterise some thermo-physical properties (thermal capacity, thermal contact, and conductivity) of a micrometric layer. The pyrometer system was developed and applied to characterise some properties of a W-layer (140 μm) on a CFC-substrate. The numerical code developed for 3-D simulation of LH of a surface with the deposited layer was applied to simulate the experimental heating temperatures. The experimental and simulation results were compared. W-layer characterisation was performed by fitting the experimental and theoretical heating temperatures.     

Tritium permeation in EUROFER in EXOTIC and LIBRETTO irradiation experiments

The reduced activation martensitic steel (RAFM) EUROFER is foreseen as a structural material in test breeder module (TBM) in ITER and breeder blanket in DEMO design. In a number of irradiation experiments conducted in high flux reactor (HFR) in Petten EUROFER was used as a containment wall of the breeder material, through which tritium permeation was monitored on line. Thus in EXOTIC-9/1 (EXtraction Of Tritium In Ceramics) experiment where Li₂TiO₃ pebbles were the breeder material, EUROFER was irradiated up to 1.3 dpa at 340–580 °C. In LIBRETTO experiments (LIBRETTO-4/1, -4/2 and -5) the breeder material was lead lithium eutectic which was in direct contact with the EUROFER containment wall. The neutron damage in steel achieved in the LIBRETTO experiments varied from 2 to 3.5 dpa. The irradiation temperature was 350 °C (LIBRETTO-4/1), 550 °C (LIBRETTO-4/2), and 300–500 °C (LIBRETTO-5).Tritium permeability was studied by varying the irradiation temperature and hydrogen concentration in the purge gas. From the analysis of the temperature transients performed in all four experiments yielded the tritium diffusion coefficients were derived, which appear to be factor ten lower than the literature data obtained in the gas driven permeation experiments.     

Novel granulation process of beryllide as advanced neutron multipliers

Advanced neutron multipliers with low swelling and high stability at high temperatures are desired for pebble bed blankets of demonstration fusion power (DEMO) reactors. Beryllium intermetallic compounds (beryllides) are the most promising advanced neutron multipliers. In order to fabricate the beryllide pebbles, beryllide with shapes of block and/or rod is necessary when a melting granulation process is applied such as a rotating electrode method. A plasma sintering method has been proposed as new technique which uses a non conventional consolidation process. It was clarified that the beryllide could be simultaneously synthesized and jointed by the plasma sintering method in the insert material region between two beryllide blocks. Beryllide rod of Be₁₂Ti with 10 mm in diameter and 60 mm in length has been successfully fabricated by the plasma sintering method. Using this plasma-sintered beryllide rod, fabrication of prototype beryllide pebble was performed by a rotating electrode method as one of the melting methods. The prototype pebbles of Be₁₂Ti with 1 mm in average diameter were successfully fabricated.     

An advance process of aluminum rich coating as tritium permeation barrier on 321 steel workpiece

We have proposed an advance three-step process, Al-electroplating in ionic liquid followed by heat treating and selectively oxidation, preparing aluminum rich coating as tritium permeation barrier (TPB). In present work, the advance process was applied to 321 steel workpieces. In the Al-electroplating, pieces were coated by galvanostatic electrodeposition at 20 mA/cm² in aluminum chloride (AlCl₃)–1-ethyl-3-methylimidazolium chloride (EMIC) ionic liquid. The Al coating on those pieces all displayed attractive brightness and well adhered to surface of pieces. Within the aluminizing time from 1 to 30 h, a series of experiments were carried out to aluminize 321 steel pieces with Al 20 μm coating at 700 °C. After heat treated for 8 h, a 30 μm thick aluminized coating on piece appeared homogeneous, free of porosity, and mainly consisted of (Fe, Cr, Ni)Al₂, and then was selectively oxidized in argon gas at 700 °C for 50 h to form Al₂O₃ scale. The finally fabricated aluminum rich coating, without any visible defects, had a double-layered structure consisting of an outer γ-Al₂O₃ layer with thickness of 0.2 μm and inner (Fe, Cr, Ni)Al/(Fe, Cr, Ni)₃Al layer of 50 μm thickness. The deuterium permeation reduction factor, PRF, of piece (Φ 80 × 2, L 150 mm) with such coating increased by 2 orders of magnitude at 600–727 °C. The reproducibility of the process was also showed.     

Inactive commissioning of a micro channel catalytic reactor for highly tritiated water production in the CAPER facility of TLK

In future DT fusion machines, several events will generate highly tritiated water (HTW). Among potential techniques for HTW processing, isotopic swamping in a catalytic membrane reactor (PERMCAT) appears promising. The experimental demonstration of PERMCAT for HTW processing has started in the CAPER facility at the Tritium Laboratory of Karlsruhe (TLK). Without any HTW source, such water has to be produced on purpose.Catalytic HT oxidation would ensure clean operation but could be critical for operation due to possible occurrence of explosive mixture. A tritium compatible micro-channel catalytic reactor (μCCR) has been designed and manufactured to produce up to 10 mL min^?1 of HTW with very high specific tritium activity (stoichiometric DTO: 5.2 × 10¹⁶ Bq kg^?1). Prior to its integration in CAPER for tritium operation, this reactor has been commissioned at different feed flow rates, gas composition (air or Helium), and temperature. The results demonstrate the good performances of the μCCR in producing water.The combination of the μCCR with the O₂ sensor represents a reliable system able to produce HTW in a safe way and without radioactive waste. Accordingly, the CAPER facility can be upgrade in order to continue the R&D activity on HTW processing with PERMCAT.     

Development of fabrication technologies for advanced breeding functional materials For DEMO reactors

Demonstration power reactors (DEMOs) require advanced tritium breeders and neutron multipliers that have high stability at high temperatures. Lithium titanate (Li₂TiO₃) is one of the most promising candidates among tritium breeders because of its tritium release characteristics. Li₂TiO₃ with additional Li (Li_2+xTiO_3+y) has increased stability in a reducing atmosphere at high temperatures. In this work, Li_2+xTiO_3+y pebbles were fabricated using the emulsion method, which is a sol–gel method. The raw material for the fabrication of Li_2+xTiO_3+y pebbles was synthesized from a mixture of LiOH·H₂O and H₂TiO₃ at specific ratios. The average diameter and the sphericity of the pebbles fabricated by the emulsion method were 1.40 mm and 1.02, respectively. In addition, beryllium (Be) intermetallic compounds (beryllides) are promising materials for advanced neutron multipliers. The results of the trial fabrications in this work showed that beryllides of Be–Ti and Be–V intermetallics could be synthesized using the plasma sintering method.     

A Fusion Nuclear Science Facility for a fast-track path to DEMO

An accelerated fusion energy development program, a “fast-track” approach, requires proceeding with a nuclear and materials testing program in parallel with research on burning plasmas, ITER. A Fusion Nuclear Science Facility (FNSF) would address many of the key issues that need to be addressed prior to DEMO, including breeding tritium and completing the fuel cycle, qualifying nuclear materials for high fluence, developing suitable materials for the plasma-boundary interface, and demonstrating power extraction. The Advanced Tokamak (AT) is a strong candidate for an FNSF as a consequence of its mature physics base, capability to address the key issues, and the direct relevance to an attractive target power plant. The standard aspect ratio provides space for a solenoid, assuring robust plasma current initiation, and for an inboard blanket, assuring robust tritium breeding ratio (TBR) >1 for FNSF tritium self-sufficiency and building of inventory needed to start up DEMO. An example design point gives a moderate sized Cu-coil device with R/a = 2.7 m/0.77 m, κ = 2.3, B_T = 5.4 T, I_P = 6.6 MA, β_N = 2.75, P_fus = 127 MW. The modest bootstrap fraction of ƒ_BS = 0.55 provides an opportunity to develop steady state with sufficient current drive for adequate control. Proceeding with a FNSF in parallel with ITER provides a strong basis to begin construction of DEMO upon the achievement of Q ∼ 10 in ITER.     

Effect of oxygen on the corrosion behavior of SS316L in a buoyancy driven Pb–17Li loop

Corrosion behavior of SS316L in lead–lithium eutectic in the presence of oxygen was investigated in a thermal convection loop for 1000 h of exposure. At a thermal gradient of 100 K, a 20 μm deep ferrite layer was formed on the exposed surface. Introduction of oxygen resulted in a substantially high chromium depletion from the steel matrix. EPMA profiles revealed the presence of chromium enriched, lead free layer over the surface facing liquid lead–lithium. XRD data confirmed the presence of LiCrO₂ and Cr₂O₃ in this layer. It is expected that this layer at the interface can act as a passive boundary and thus prevent continued corrosion by liquid metal.     

Neutronic model of a mirror based fusion–fission hybrid for the incineration of the transuranic elements from spent nuclear fuel and energy amplification

The Georgia Institute of Technology has developed several design concepts of tokamak based fusion–fission hybrids for the incineration of the transuranic elements of spent nuclear fuel from Light-Water-Reactors. The present paper presents a model of a mirror hybrid. Concerning its main operation parameters it is in several aspects analogous to the first tokamak based version of a “fusion transmutation of waste reactor”. It was designed for a criticality k_eff ≤ 0.95 in normal operation state. Results of neutron transport calculations carried out with the MCNP5 code and with the JEFF-3.1 nuclear data library show that the hybrid generates a fission power of 3 GW_th requiring a fusion power between 35 and 75 MW, has a tritium breeding ratio per cycle of TBR_cycle = 1.9 and a first wall lifetime of 12–16 cycles of 311 effective full power days. Its total energy amplification factor was roughly estimated at 2.1. Special calculations showed that the blanket remains in a deep subcritical state in case of accidents causing partial or total voiding of the lead–bismuth eutectic coolant. Aiming at the reduction of the required fusion power, a near-term hybrid option was identified which is operated at higher criticality k_eff ≤ 0.97 and produces less fission power of 1.5 GW_th. Its main performance parameters turn out substantially better.     

Study of ceramic pebble beds in Post Irradiation Examination of the Pebble Bed Assemblies irradiation experiment

The irradiation experiment Pebble Bed Assemblies (PBA) consists of four mock-up representations (test elements) of the EU Helium Cooled Pebble Bed (HCPB) concept. The four test elements contain a ceramic breeder pebble bed sandwiched between two beryllium pebble beds and are regarded as one of the first DEMO representative HCPB blanket irradiation tests, with respect to temperatures and power densities. The design value of the PBA were to irradiate pebble beds at a power density of 20–26 W/cc in the ceramic breeder, to a maximum temperature of 800 °C.Two test elements contain lithium orthosilicate pebbles (Li₄SiO₄; FZK/KIT) and were irradiated with target temperatures of 600 and 800 °C, respectively. The other test elements have lithium metatitanate (Li₂TiO₃; CEA) with different grain sizes and were both irradiated with a target temperature of 800 °C. The PBA have been irradiated for 294 Full Power Days (12 cycles) in the High Flux Reactor (HFR) in Petten to a total neutron dose of 2–3 dpa in Eurofer, and an estimated (total) lithium burnup of 2–3% in the ceramic pebbles.This work presents results of Post Irradiation Examinations (PIE) on the four HCPB test elements. Using e.g. SEM, the evolution of compressed pebble beds and pebble interactions like swelling, creep, sintering, etc., under irradiation and thermal loads are studied for the candidate pebble materials Li₂TiO₃ and Li₄SiO₄. (Chemical) interactions between ceramic pebbles and Eurofer (e.g. chrome diffusion) are observed. Looking at different sections of the pebble beds, correlations between temperatures and thermal–mechanical behaviour are clearly observed.     

Biological studies using mammalian cell lines and the current status of the microbeam irradiation system,SPICE

The development of SPICE (single-particle irradiation system to cell), a microbeam irradiation system, has been completed at the National Institute of Radiological Sciences (NIRS). The beam size has been improved to approximately 5 μm in diameter, and the cell targeting system can irradiate up to 400–500 cells per minute. Two cell dishes have been specially designed: one a Si₃N₄ plate (2.5 mm × 2.5 mm area with 1 μm thickness) supported by a 7.5 mm × 7.5 mm frame of 200 μm thickness, and the other a Mylar film stretched by pressing with a metal ring. Both dish types may be placed on a voice coil stage equipped on the cell targeting system, which includes a fluorescent microscope and a CCD camera for capturing cell images. This microscope system captures images of dyed cell nuclei, computes the location coordinates of individual cells, and synchronizes this with the voice coil motor stage and single-particle irradiation system consisting of a scintillation counter and a beam deflector. Irradiation of selected cells with a programmable number of protons is now automatable. We employed the simultaneous detection method for visualizing the position of mammalian cells and proton traversal through CR-39 to determine whether the targeted cells are actually irradiated. An immuno-assay was also performed against γ-H2AX, to confirm the induction of DNA double-strand breaks in the target cells.     

Experimental set-up for time resolved small angle X-ray scattering studies of nanoparticles formation using a free-jet micromixer

Recently, we have designed, fabricated and tested a free-jet micromixer for time resolved small angle X-ray scattering (SAXS) studies of nanoparticles formation in the <100 μs time range. The microjet has a diameter of 25 μm and a time of first accessible measurement of 75 μs has been obtained. This result can still be improved. In this communication, we present a method to estimate whether a given chemical or biological reaction can be investigated with the micromixer, and to optimize the beam size for the measurement at the chosen SAXS beamline. Moreover, we describe a system based on stereoscopic imaging which allows the alignment of the jet with the X-ray beam with a precision of 20 μm. The proposed experimental procedures have been successfully employed to observe the formation of calcium carbonate (CaCO₃) nanoparticles from the reaction of sodium carbonate (Na₂CO₃) and calcium chloride (CaCl₂). The induction time has been estimated in the order of 200 μs and the determined radius of the particles is about 14 nm.     

Microstructure and properties of tungsten–samarium oxide composite prepared by a novel wet chemical method and spark plasma sintering

W–1 wt% Sm₂O₃ powders doped with highly uniform Sm₂O₃ were successfully synthesized by a novel wet chemical method followed by hydrogen reduction. The powders were consolidated by spark plasma sintering (SPS) at 1800 °C to suppress grain growth during sintering. The FE-SEM and HRTEM analysis, tensile test and thermal conductivity measurements were used to characterize these samples. The grain size, relative density of the bulk samples fabricated by SPS sintering were 4 μm and 97.8%, respectively. The tensile strength values of Sm₂O₃/W samples were higher than those of pure W samples. As the temperature rises from 25 to 800 °C, the thermal conductivity of pure W and W–1 wt% Sm₂O₃ composites decreased with the same trend and the thermal conductivity of both samples was above 160 W/m K at room temperature.     

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.