Liquid metal MHD studies with non-magnetic and ferro-magnetic structural material

In most of the liquid metal MHD experiments reported in the literature to study liquid breeder blanket performance, SS316/SS304 grade steels are used as the structural material which is non-magnetic. On the other hand, the structural material for fusion blanket systems has been proposed to be ferritic martensitic grade steel (FMS) which is ferromagnetic in nature. In the recent experimental campaign, liquid metal MHD experiments have been carried out with two identical test sections: one made of SS316L (non-magnetic) and another with SS430 (ferromagnetic), to compare the effect of structural materials on MHD phenomena for various magnetic fields (up to 4 T). The maximum Hartmann number and interaction number are 1047 and 300, respectively.Each test section consists of square channel (25 mm × 25 mm) cross-section with two U bends, with inlet and outlet at the middle portion of two horizontal legs, respectively. Pb–Li enters into the test section through a square duct and distributed into two parallel paths through a partition plate. In each parallel path, it travels ∼0.28 m length in plane perpendicular to the magnetic field and faces two 90° bends before coming out of the test section through a single square duct. The wall electrical potential and MHD pressure drop across the test sections are compared under identical experimental conditions. Similar MHD behavior is observed with both the test section at higher value of the magnetic field (>2 T).     

Numerical investigation of liquid metal magnetohydrodynamic flow in multilayer flow channel inserts

In the high temperature liquid metal blanket of fusion-based hydrogen production reactor (named FDS-III), there is a remarkable feature that the multilayer flow channel inserts (MFCI) as function component are put into the breeding zone. The low thermal conductivity of MFCI can prevent the internal PbLi's heat conduct to the outside. So the outlet temperature can achieve high temperature around 1000 °C for high efficient production of hydrogen. However, the flow of liquid metal meandering through the MFCI will cause complex magnetohydrodynamic (MHD) effect under the strong fusion magnetic field. Liquid metal MHD effect is a key issue which should be concerned in high temperature breeder blanket (HTL). In this work, a numerical study was carried out to investigate the MHD effect of liquid metal PbLi in the MFCI. The MHD flows with typical modified geometry of the HTL MFCI were considered. The characteristics of flow and induced current fields were analyzed, and the pressure drop was evaluated. It also can be seen that the conductivity of the MFCI will have great impact on liquid metal flow's current and velocity distributions.     

Numerical simulation of a blanket cooling system for fusion reactor based on PWR conditions

The simulations of a blanket cooling system were presented to address the choice of cooling channel geometry and coolant input data which are related to blanket engineering implementation. This work was performed using computer aided design (CAD) and computational fluid dynamics (CFD) technology. Simulations were carried out for the blanket module with a size of 0.6 m × 0.45 m in toroidal plane, and the nuclear heat was applied on the cooling system at P_n (neutron wall load) of 5 MW/m². The structure factors and input data of hydraulics were investigated to explore the optimal parameters to match the PWR condition. It was found that the inlet velocity of first wall (FW) channel should be within the range of 2.48–3.34 m/s. As a result, the temperature rise (TR) of the coolant in the FW channel would be 24–25 K. This leads to the remaining space for TR within the range of 15 K in the piping circuits. It also indicated that the FW plays an important role in TR (reaches 60% of the whole cooling system) due to its high level of P_n and heat flux in the zones. It was predicted that the nuclear heat inside blanket module could be removed completely by the piping circuits with an acceptable pipe bore and the related input data. Finally, a possible design range of cooling parameters was proposed in view of engineering feasibility and blanket neutronics design.     

Design and fabrication of a high-temperature helium regenerator

Refractory metallic foams can increase heat transfer efficiency in gas-to-gas and liquid metal-to-gas heat exchangers by providing an extended surface area for better convection, i.e. conduction into the foam ligaments providing a “fin-effect,” and by disruption of the thermal boundary layer near the hot wall and ligaments by turbulence promotion. In this article, we describe the design of a high-temperature refractory regenerator (closed-loop recuperator) using computational fluid dynamics (CFD) modeling of actual foam geometries obtained through computerized micro-tomography. The article outlines the design procedure from geometry import through meshing and thermo-mechanical analysis and discusses the challenges of fabrication using pure molybdenum and TZM. The foam core regenerator is more easily fabricated, less expensive and performs better than refractory flat plate-type heat exchangers. The regenerator can operate with a maximum hot leg inlet temperature of 900 °C and transfer 180 kW to the cold leg using 100 g/s helium at 4 MPa. Future high heat flux experiments on helium-cooled plasma facing components will utilize the high temperature and high pressure capabilities of this unique regenerator. Similar components will be required to adapt fusion power reactors to high-efficiency Brayton power conversion systems and enable operation of advanced divertor and blanket systems.     

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.     

Enhanced arrangement for recuperators in supercritical CO2 Brayton power cycle for energy conversion in fusion reactors

A domestic research program called TECNO_FUS was launched in Spain in 2009 to support technological developments related to a dual coolant breeding blanket concept for fusion reactors. This concept of blanket uses Helium (300 °C/400 °C) to cool part of it and a liquid metal (480 °C/700 °C) to cool the rest; it also includes high temperature (700 °C/800 °C) and medium temperature (566 °C/700 °C) Helium cooling circuits for divertor. This paper proposes a new layout of the classical recompression supercritical CO₂ Brayton cycle which replaces one of the recuperators (the one with the highest temperature) by another which by-passes the low temperature blanket source. This arrangement allows reaching high turbine inlet temperatures (around 600 °C) with medium pressures (around 225 bar) and achieving high cycle efficiencies (close to 46.5%). So, the proposed cycle reveals as a promising design because it integrates all the available thermal sources in a compact layout achieving high efficiencies with the usual parameters prescribed in classical recompression supercritical CO₂ Brayton cycles.     

Gravitational flow of a thin film of liquid metal in a strong magnetic field

The influence of a poloidal magnetic field of the spherical Tokamak on super thin (h ≈ 0.1 mm) film flow of liquid metal driven by gravity over the surface of the cooled divertor plate is addressed. The experimental setup developed at the Institute of Physics, University of Latvia (IPUL) is described, which makes it possible to drive and visualize such liquid metal flows in the solenoid of the superconducting magnet “Magdalena”. As applied to the above setup, the magnetic field effect on the operation of the capillary system of liquid metal flow distribution (CSFD) is evaluated by using molten metal (lithium or eutectic InGaSn alloy) with a very small linear flowrate q ≤ 1 mm²/s, spread uniformly across the substrate. The magnetic field effect on the main parameters of the fully developed film flow is estimated for the above-mentioned liquid metals.An approximation technique has been proposed to calculate the development of the gravitational film flow. A non-linear differential second order equation has been derived, which describes the variation of the film flow thickness over the substrate length versus the flowrate q, magnetic field B and the substrate sloping α.Results of InGaSn film flow observations in a strong (B = 4 T) poloidal magnetic field are presented. Analysis of the video records evidences of experimental realization of a stable stationary film flow at width-uniform supply of InGaSn.     

Monte Carlo calculations of the nuclear effects of certain fluids in a hybrid reactor

This study analyzes the effects of certain heavy-metal-salt fluids on nuclear parameters in a fusion–fission hybrid reactor. Calculated parameters include the tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate, fission reaction rate, and fissile fuel breeding in the reactor's liquid first wall, blanket, and shield zones; gas production rates in the structural material of the reactor were calculated, as well. The fluid mixtures consisted of 93–85% Li₂₀Sn₈₀ + 5% SFG-PuO₂ and 2–10% UO₂, 93–85% Li₂₀Sn₈₀ + 5% SFG-PuO₂ and 2–10% NpO₂, and 93–85% Li₂₀Sn₈₀ + 5% SFG-PuO₂ and 2–10% UCO. The fluids were used in the liquid first wall, blanket, and shield zones of a fusion–fission hybrid reactor system. A 3 cm wide beryllium (Be) zone was used for neutron multiplier between the liquid first wall and the blanket. The structural material used was 9Cr2WVTa ferritic steel, measuring 4 cm in width. Three-dimensional analyses were performed using the Monte Carlo code MCNPX-2.7.0 and the ENDF/B-VII.0 nuclear data library.     

Preliminary design of the ITER ECH Upper Launcher

The design of the ITER electron cyclotron launchers recently reached the preliminary design level - the last major milestone before design finalization. The ITER ECH system contains 24 installed gyrotrons providing a maximum ECH injected power of 20 MW through transmission lines towards the tokamak. There are two EC launcher types both using a front steering mirror; one equatorial launcher (EL) for plasma heating and four upper launchers (UL) for plasma mode stabilization (neoclassical tearing modes and the sawtooth instability). A wide steering angle range of the ULs allows focusing of the beam on magnetic islands which are expected on the rational magnetic flux surfaces q = 1 (sawtooth instability), q = 3/2 and q = 2 (NTMs).In this paper the preliminary design of the ITER ECH UL is presented, including the optical system and the structural components. Highlights of the design include the torus CVD-diamond windows, the frictionless, front steering mechanism and the plasma facing blanket shield module (BSM). Numerical simulations as well as prototype tests are used to verify the design     

Model validation of GAMMA code with heat transfer experiment for KO TBM in ITER

By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He cooled molten lithium (HCML) test blanket module (TBM) for testing in ITER. A performance analysis for the thermal–hydraulics and a safety analysis for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs multicomponent mixture analysis), which was developed by the gas cooled reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM first wall (FW) mock-up made from the same material as the KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 1.1, 1.9 and 2.9 MPa, and under various ranges of flow rate from 0.0105 to 0.0407 kg/s with a constant wall temperature condition. In the present study, a thermal–hydraulic test was performed with the newly constructed helium supplying system, in which the design pressure and temperature were 9 MPa and 500 °C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 3 MPa pressure, 30 °C inlet temperature and 70 m/s helium velocity, which are almost same conditions of the KO TBM FW. One side of the mock-up was heated with a constant heat flux of 0.3–0.5 MW/m² using a graphite heating system, KoHLT-2 (Korea heat load test facility-2). Because the comparison result between CFX 11 and GAMMA showed a difference tendency, the modification of heat transfer correlation included in GAMMA was performed. And the modified GAMMA showed the strong parity with CFX 11 calculation results.     

The ARIES-AT advanced tokamak,Advanced technology fusion power plant

The ARIES-AT study was initiated to assess the potential of high-performance tokamak plasmas together with advanced technology in a fusion power plant and to identifying physics and technology areas with the highest leverage for achieving attractive and competitive fusion power in order to guide fusion R&D. The 1000-MWe ARIES-AT design has a major radius of 5.2 m, a minor radius of 1.3 m, a toroidal β of 9.2% (β_N = 5.4) and an on-axis field of 5.6 T. The plasma current is 13 MA and the current-drive power is 35 MW. The ARIES-AT design uses the same physics basis as ARIES-RS, a reversed-shear plasma. A distinct difference between ARIES-RS and ARIES-AT plasmas is the higher plasma elongation of ARIES-AT (κ_x = 2.2) which is the result of a “thinner” blanket leading to a large increase in plasma β to 9.2% (compared to 5% for ARIES-RS) with only a slightly higher β_N. ARIES-AT blanket is a simple, low-pressure design consisting of SiC composite boxes with a SiC insert for flow distribution that does not carry any structural load. The breeding coolant (Pb–17Li) enters the fusion core from the bottom, and cools the first wall while traveling in the poloidal direction to the top of the blanket module. The coolant then returns through the blanket channel at a low speed and is superheated to ∼1100 °C. As most of the fusion power is deposited directly into the breeding coolant, this method leads to a high coolant outlet temperature while keeping the temperature of the SiC structure as well as interface between SiC structure and Pb–17Li to about 1000 °C. This blanket is well matched to an advanced Brayton power cycle, leading to an overall thermal efficiency of ∼59%. The very low afterheat in SiC composites results in exceptional safety and waste disposal characteristics. All of the fusion core components qualify for shallow land burial under U.S. regulations (furthermore, ∼90% of components qualify as Class-A waste, the lowest level). The ARIES-AT study shows that the combination of advanced tokamak modes and advanced technology leads to an attractive fusion power plant with excellent safety and environmental characteristics and with a cost of electricity (4.7 ¢/kWh), which is competitive with those projected for other sources of energy.     

The lower hybrid current drive system for steady-state operation of the Vulcan tokamak conceptual design

The steady-state current drive system for the Vulcan tokamak concept has been designed, taking into account requirements of high field, small size, and high operational wall temperature (B₀ = 7 T, R₀ = 1.2 m, T_wall > 800 K). This lower hybrid current drive system allows steady-state operation by utilizing high field side launch, high RF source frequency (8 GHz), and dedicated current drive ports. An iterative MHD and current drive solver is used to determine the ideal launching spectra and location to assure strong single pass absorption. It is found that with nominal Vulcan operational parameters (n_e ≈ 4 × 10²⁰ m^?3, T_e ≈ 2.8 keV, I_p = 1.7 MA, P_LHCD = 19.8 MW) bootstrap currents of ～70% and lower hybrid current drive efficiencies of 1.16 × 10¹⁹ A W m^?2 could be achieved. The optimized solution yielded advanced tokamak profiles with q values on-axis above 2. A conceptual design of the system is presented, which takes into account space, power, cooling, and launched spectrum requirements. The system is found to be compatible with the vacuum vessel design and requires cooling power of <1 MW per waveguide bundle.     

Development of water-cooled solid breeder test blanket module in JAEA

One of the most important missions of ITER is to provide a test bed for breeding blanket modules, which are called as test blanket module (TBM). JAEA has been extensively developing a water-cooled solid breeder test blanket module (WCSB TBM) for ITER. JAEA developed fabrication technology of F82H rectangular cooling tubes, and has successfully fabricated the near-full scale first wall mock-up of WCSB TBM by hot isostatic press (HIP) technique, which is fully made of F82H. The mock-up has been high-heat flux tested in the DATS facility in JAEA, which is an ion beam test facility. The inlet temperature of the cooling water is about 280 °C with 15 MPa, which is almost the same as the WCSB TBM design conditions. The mock-up has endured a heat load of 0.5 MW/m², 30 s for 80 thermal cycles. Neither hot spots nor thermal degradation have been observed.     

A method to produce lithium pellets for fueling and ELM pacing in NSTX-U

A device for producing small, high frequency spherical droplets or pellets for lithium or other liquid metals has been developed and could aid in the controlled excitation or pacing of edge-localized plasma modes (ELMs). The Liquid Lithium/metal Pellet Injector (LLPI) could also be used to replenish lithium coatings of plasma-facing components (PFCs) during a plasma discharge. With NSTX-U having longer pulse lengths (up to 5 s), it is desirable to be able to inject lithium during the discharge to maintain the beneficial effects. Using a nozzle injector design and a surrogate to lithium, Wood's metal, the LLPI has achieved droplet diameters between 0.6 mm < d_drop < 1 mm in diameter and frequencies up to 1.5 kHz with argon gas driving the formation. This paper presents the LLPI being developed with initial results mainly using Wood's metal and some lithium, using high pressure argon to force the liquid lithium through the nozzle.     

Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency (TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%.     

Preparation and investigation of aluminized coating and subsequent heat treatment on 9Cr–1Mo Grade 91 steel

Iron aluminide inner coating with alumina top layer is being considered as a potential solution for tritium permeation barrier and mitigating MHD pressure drop for liquid metal blanket concepts in the fusion reactor systems. Hot-dip aluminizing with subsequent heat treatment seems to offer a good possibility to produce aluminized coating with alumina top layer. 9Cr–1Mo Grade 91 steel samples were hot dipped in Al melt containing 2.25 wt% of Si at 750 °C for 3 min. Heat treatment was performed at 650, 750 and 950 °C for 5 h; samples were either air cooled or furnace cooled. Coatings have been evaluated by SEM, EDX, X-ray diffraction, microhardness, scratch adhesion and Raman spectroscopy. The thickness of the layers and phases formed were influenced by the heat treatment adopted. Fe₂Al₅ was the major phase present in the samples heat treated at 650/750 °C, whereas FeAl and α-Fe(Al) primarily made up the outer and inner layers respectively in the samples heat treated at 950 °C. Cooling method deployed affected the hardness. Air cooled samples had comparatively higher hardness than furnace cooled samples. The scratch test showed the adhesion for the samples heat treated at 950 °C was much better as compared to the samples heat treated at 650/750 °C. Raman spectroscopy analysis showed the presence of both α-Al₂O₃ and γ-Al₂O₃ on the surface of the samples heat treated at 950 °C, while Fe₃O₄ was present in the furnace cooled sample only.     

Direct numerical simulation of turbulent mixed convection in a vertical channel in a wall-normal magnetic field

Liquid metal coolants have a significant role in the design of advanced fusion reactors. There is a need for an investigation of the thermal behavior of the liquid metal in working reactor environment, such as when fluid flow at low Prandtl number (Pr) with a buoyancy effect, is subjected to a magnetic field. In the present study, a direct numerical simulation (DNS) for a low Pr number fluid flow resulting in turbulent heat transfer with buoyancy effect under a magnetic field has been carried out between two vertical plates kept at different temperatures. In this simulation, the values of the Hartmann number (Ha) were 0 and 6, Pr number was 0.06 and Grashof numbers were 6.4 × 10⁵, 9.6 × 10⁵, and 1.6 × 10⁶. The turbulent quantities of the parameters such as the mean temperature, turbulent heat flux, and temperature variance were obtained by direct numerical simulation (DNS). The Reynolds number (Re) for channel flow based on friction velocity averaged by both walls, viscosity, and channel half-width was set to be constant as Re_τ* = 150. A uniform magnetic field was applied in a direction perpendicular to the walls of the channel. The profiles of mean velocity and velocity fluctuations became asymmetric, and the tendency was enhanced with the increasing buoyancy effect. However, by the application of a magnetic field the tendency decreased. In other words, thermal transport between the walls became weak due to the magnetic effect.     

Creep-to-rupture of the steel P92 at 650 °C in oxygen-controlled stagnant lead in comparison to air

Creep-to-rupture experiments were performed on 9%-Cr ferritic–martensitic steel P92 in the CRISLA facility. The specimens of P92 were examined at 650 °C and static tensile stress between 75 and 325 MPa in both stagnant lead with 10^?6 mass% dissolved oxygen and air. The steel showed an insignificant difference in time-to-rupture, t_R, and ductile fracture in both environments at >100 MPa, corresponding to t_R < 3,442 h. At 75 MPa in Pb (t_R = 13,090 h), the steel, however, featured purely brittle fracture pointing to liquid metal embrittlement. Structural changes in the steel and surface oxidation in the different environments were studied using metallographic techniques. The Laves phase that forms during thermal aging at 650 °C was found along prior austenite grain boundaries and martensite laths already after relatively short testing time, along with chromium carbides that are already present in the as-received condition of the steel.     

Fabrication and high heat flux test of large mockups for ITER first wall semi-prototype

The main topic of an ITER blanket first wall procurement is to qualify whether each party has the key technology needed for the fabrication and joining of first wall components. A semi-prototype qualification project will be released requiring that the single components of a full-scale first wall must be fabricated and successfully pass high heat flux tests using a hypervapotron cooling channel. In this work, various mockup types have been modeled and fabricated to develop the joining technology for a semi-prototype. The semi-prototype, which has three double-fingered panels, is a scaled-down component of a full-size first wall. The standard or slit mockups with a 80 mm × 80 mm single beryllium tile joined to a CuCrZr heat sink were fabricated to qualify our HIP (Hot Isostatic Pressing) technology for the joining of semi-prototype. These standard mockups were installed to perform a high heat flux test in the Korea heat load test facility (KoHLT). For a preliminary test of a semi-prototype, thermo-hydraulic mockups of 710 mm × 100 mm were designed and fabricated to verify the Cu/SS cooling performance, such as hypervapotron. For the high heat flux test in our KoHLT facility, the normal cycle is based on an expected heat flux of 300 s in accordance with the ITER qualification specifications. These tests will be performed to qualify the joining technologies, which is required for an ITER blanket first wall and a semi-prototype.     

Verification test results of a cutting technique for the ITER blanket cooling pipes

For replacement of the first wall (FW) of the international thermonuclear experimental reactor (ITER), cutting and welding tools for the cooling pipes must be able to access a pipe from the surface side of the FW and cut/weld the pipe from the inside the cooling pipe (inner diameter: 42.72 mm, thickness: 2.77 mm). The cutting tool for the pipe end is required to cut a flat plate circularly from the surface side of the FW (cutting diameter: approximately 44 mm, plate thickness: 5 mm). To determine the specifications for both the tools and the blanket hydraulic connections, the ITER Organization (IO) and the Japan Domestic Agency (JADA) conducted research and development activities regarding the FW replacement. This paper describes the current status of the development of cutting tools for the cooling pipe connection.