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.     

ENHS reactor power level enhancement possibilities

The ENHS thermal hydraulic optimization code was modified and applied to search for the maximum attainable power from a wide range of ENHS design options subjected to the following constraints: maximum permissible hot channel coolant outlet temperature of 600 °C, clad inner temperature of 650 °C and primary coolant temperature rise of either 150 °C or 90% of the theoretical limit for accelerated corrosion rate. The T–H optimization variables include the intermediate heat exchanger number of channels, channel width and elevation; diameter of the riser and diameter of a flow-splitting shroud in the riser. It was found possible to increase the attainable power from the nominal 125 MW_th up to 311 MW_th for the reference core, 400 MW_th for a reference-like core having equilibrium composition fuel and 372 MW_th for a flattened power core with 9 plutonium concentration zones. A power level exceeding 400 MW_th may be achieved by flattening the power distribution of the equilibrium core or using nitride fuel with enriched nitrogen rather than metallic fuel. With forced circulation it is possible to operate the flattened power core at up to 532 MW_th corresponding to 223 MW_e.     

Nuclear performance assessment of ARIES-AT

The transition to a new ARIES design always involves a significant change in the engineering system with emphasis on high performance. Compared to ARIES-RS, numerous improvements were noted in ARIES-AT, ranging from physics improvements to a focus on advanced engineering for an economically and environmentally attractive source of energy. During the course of the ARIES-AT study, we closely monitored the key nuclear parameters and called for measures to enhance the engineering and physics aspects of the design. Optimization of components’ constituents, characterization of the radiation environment, and meeting the ARIES-AT specific design needs were also given considerable attention. A key engineering aspect of ARIES-AT is its compactness and the high-conversion efficiency (60%) of the LiPb/SiC blanket that is capable of performing at high temperature (1000 °C). Certain features of the nuclear activity were focused on areas unique to ARIES-AT, including breeding potential of the LiPb/SiC blanket containing tungsten stabilizing shells, high-performance shielding components to protect the high-temperature superconducting magnet, and compact radial builds to minimize the volume of solid waste requiring near-surface geological burial. The final design satisfies the top-level requirements and no serious nuclear issues have been identified for ARIES-AT. A salient design feature is the significant reduction in the ARIES-AT radwaste volume relative to precedent designs developed since the inception of the ARIES project in the late 1980s. Design measures such as the high-power density, well optimized shield, and blanket segmentation with extended service lifetime have all contributed to the waste minimization.     

Effects of heat treatment and cooling rate on microstructural and mechanical characteristics of aluminide coatings on MANET steel

R & D studies concerning Pb-17Li blankets for DEMO suggest the use of coatings either as tritium permeation barriers for the water cooled blanket or as electrical insulators for the self cooled blanket. The production of coatings resistant to the thermal and mechanical stresses typical of the blanket operating conditions is the target. The present work describes the results of an experimental programme aimed at evaluating the effects of heat treatment and cooling rate on the microstructural and mechanical characteristics of aluminide coatings deposited on MANET steel (DIN 1.4914) by the hot-dipping process. The temperature of the post-deposition heat treatment that allows obtaining structures like FeAl and -Fe(Al), instead of the more brittle Fe_xAl_y intermetallics, has been identified. The coatings have been characterized by optical metallography, microhardness measurements, scanning electron microscopy (SEM), energy dispersive X-ray analysis (EDS) and X-ray diffraction on the surface.     

The solubility of aluminium in Pb-17Li

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

Study of steam, helium and supercritical CO2 turbine power generations in prototype fusion power reactor

Power generation systems such as steam turbine cycle, helium turbine cycle and supercritical CO₂ (S-CO₂) turbine cycle are examined for the prototype nuclear fusion reactor. Their achievable cycle thermal efficiencies are revealed to be 40%, 34% and 42% levels for the heat source outlet coolant temperature of 480 °C, respectively, if no other restriction is imposed. In the current technology, however, low temperature divertor heat source is included. In this actual case, the steam turbine system and the S-CO₂ turbine system were compared in the light of cycle efficiency and plant cost. The values of cycle efficiency were 37.7% and 36.4% for the steam cycle and S-CO₂ cycle, respectively. The construction cost was estimated by means of component volume. The volume became 16,590 m³ and 7240 m³ for the steam turbine system and S-CO₂ turbine system, respectively. In addition, separation of permeated tritium from the coolant is much easier in S-CO₂ than in H₂O. Therefore, the S-CO₂ turbine system is recommended to the fusion reactor system than the steam turbine system.     

Liquid metal magnetohydrodynamic flows in manifolds of dual coolant lead lithium blankets

An attractive blanket concept for a fusion reactor is the dual coolant lead lithium (DCLL) blanket where reduced activation steel is used as structural material and a lead lithium alloy serves both to produce tritium and to remove the heat in the breeder zone. Helium is employed to cool the first wall and the blanket structure.Some critical issues for the feasibility of this blanket concept are related to complex induced electric currents and 3D magnetohydrodynamic (MHD) phenomena that occur in distributing and collecting liquid metal manifolds. They can result in large pressure drop and undesirable flow imbalance in parallel poloidal ducts forming blanket modules.In the present paper liquid metal MHD flows are studied for different design options of a DCLL blanket manifold with the aim of identifying possible sources of flow imbalance and to predict velocity and pressure distributions.     

Active radiative liquid lithium divertor concept

Developing a reactor compatible divertor has been identified as a particularly challenging technology problem for magnetic confinement fusion. Application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising Li results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor (RLLD) concept [1]. In the RLLD, Li is evaporated from the liquid lithium (LL) coated divertor strike point surface due to the intense heat flux. The evaporated Li is readily ionized by the plasma due to its low ionization energy, and the poor Li particle confinement near the divertor plate enables ionized Li ions to radiate strongly, resulting in a significant reduction in the divertor heat flux. This radiative process has the desired effect of spreading the localized divertor heat load to the rest of the divertor chamber wall surfaces, facilitating divertor heat removal. The modeling results indicated that the Li radiation can be quite strong, so that only a small amount of Li (∼a few mol/s) is needed to significantly reduce the divertor peak heat flux for typical reactor parameters. In this paper, we examine an active version of the RLLD, which we term ARLLD, where LL is injected in the upstream region of divertor. We find that the ARLLD has similar effectiveness in reducing the divertor heat flux as the RLLD, again requiring only a few mol/s of LL to significantly reduce the divertor peak heat flux for a reactor. An advantage of the ARLLD is that one can inject LL proactively even in a feedback mode to insure the divertor peak heat flux remains below an acceptable level, providing the first line of defense against excessive divertor heat loads which could result in damage to divertor PFCs. Moreover, the low confinement property of the divertor (i.e., <1 ms for Li particle confinement time) makes the ARLLD response fast enough to mitigate the effects of possible transient events such as large ELMs.     

Preliminary Analysis of Liquid Metal MHD Pressure Drop in the Blanket for the FDS

Preliminary analysis and calculation of liquid metal Li17Pb83 magnetogydrodynamic (MHD) pressure drop in the blanket for the FDS have been presented to evaluate the significance of MHD effects on the thermal-hydraulic design of the blanket.To decrease the liquid metal MHD pressure drop,Al2O3 is applied as an electronically insulated coating onto the inner surface of the ducts.The requirement for the insulated coating to reduce the additional leakage pressure drop caused by coating imperfections has been analyzed.Finally,the total liquid metal MHD pressure drop and magnetic pump power in the FDS blanket have been given.     

Temperature distribution in mixed ThO2–UO2 fuel rods located in blanket of an inertial fusion energy breeder

Investigations of neutronic analysis and temperature distribution in fuel rods located in a blanket driven ICF (Inertial Confinement Fusion) have been performed for various mixed fuels and coolants under a first wall load of 5 MW/m². The fuel rods containing ThO₂ and UO₂ mixed by various mixing methods for achieving a flat fission power density are replaced in the blanket and cooled with different coolants; natural lithium, flibe, eutectic lithium and helium for the nuclear heat transfer. It is assumed that surface temperature of the fuel rod increases linearly from 500 °C (at top) to 700 °C (at bottom) during cooling fuel zone. Neutronic and temperature distribution calculations have been performed by MCNP4B Code and HEATING7, respectively. In the blanket fueled with pure UO₂ and cooled with helium, M (fusion energy multiplication ratio) increases to 3.9 due to uranium having higher fission cross-section than thorium. The high fission energy released in this blanket, therefore, causes proportionally increasing of temperature in the fuel rods to 823 °C. However, the M is 2.00 in the blanket fueled with pure ThO₂ and cooled with eutectic lithium because of more capture reaction than fission reaction. Maximum and minumum values of TBR (tritium breeding ratio) being one of main neutronic paremeters for a fusion reactor are 1.07 and 1.45 in the helium and the natural lithium coolant blanket, respectively. These consequences bring out that the investigated reactor can produce substantial electricity in situ during breeding fissile fuel and can be self-sufficient in the tritium required for the DT fusion driver in all cases of mixed fuels and coolant types. Quasi-constant fission power density profiles in FFB (fissile fuel breeding) zone are obtained by parabolically increasing mixture fraction of UO₂ in radial and axial directions for all coolant types. Such as, in the helium coolant blanket and the case of PMF (parabolically mixed fuel), Γ (peek-to-average fission power density ratio) of the blanket is reduced to 1.1, and the maximum temperatures of the fuel rods in radial direction of the FFB zone are also quasi-constant. At the same time, in the case of PMF, for all coolant types, the temperature profiles in the radial direction of the fuel rods rise proportionally with surface temperature from the top to the bottom of fuel rods in the axial direction. In other words, for each radial temperature profile in the axial direction, temperature differences between centerline and surface of the fuel rods are quasi-constant. According to the coolant types, these temperature diffences vary between 30 and 45 °C.     

VVER反应堆燃料组件流动传热特性CFD分析

采用计算流体力学（CFD）方法对俄罗斯水-水高能反应堆（VVER）先进燃料组件（AFA）的流动传热特性进行模拟,获得了额定工况下燃料组件冷却剂流场、流动压降和温度分布等。结果表明:与内部含交混翼的格架相比,AFA燃料组件定位格架的压力损失较小;定位格架围板导向翼附近存在滞流现象,导致燃料组件外围区域冷却剂温度偏高;不同的测量管周向棒功率比K_c对燃料组件出口冷却剂温度的测量值有较大影响。该分析结果可为核电厂堆芯温升预警值ΔT_t的设定提供参考。      

ARIES-AT safety design and analysis

ARIES-AT is a 1000 MWe conceptual fusion power plant design with a very low projected cost of electricity. The design contains many innovative features to improve both the physics and engineering performance of the system. From the safety and environmental perspective, there is greater depth to the overall analysis than in past ARIES studies. For ARIES-AT, the overall spectrum of off-normal events to be examined has been broadened. They include conventional loss of coolant and loss of flow events, an ex-vessel loss of coolant, and in-vessel off-normal events that mobilize in-vessel inventories (e.g., tritium and tokamak dust) and bypass primary confinement such as a loss of vacuum and an in-vessel loss of coolant with bypass. This broader examination of accidents improves the robustness of the design from the safety perspective and gives additional confidence that the facility can meet the no-evacuation requirement under average weather conditions. We also provide a systematic assessment of the design to address key safety functions such as confinement, decay heat removal, and chemical energy control. In the area of waste management, both the volume of the component and its hazard are used to classify the waste. In comparison to previous ARIES designs, the overall waste volume is less because of the compact design.     

Fusion-Driven Sub-Critical Dual-Cooled Waste Transmutation Blanket： Design and Analysis

The Fusion-Driven Sub-critical System (FDS) is one of the Chinese programs to be further developed for fusion application. Its Dual-cooled Waste Transmutation Blanket (DWTB), as one the most important part of the FDS is cooled by helium and liquid metal, and have the features of safety, tritium self-sustaining, high efficiency and feasibility. Its conceptual design has been finished. This paper is mainly involved with the basic structure design and thermal-hydraulics analysis of DWTB. On the basis of a three-dimensional (3-D) model of radial-toroidal sections of the segment box, thermal temperature gradients and structure analysis made with a comprehensive finite element method (FEM) have been performed with the computer code ANSYS5.7 and computational fluid dynamic finite element codes. The analysis refers to the steady-state operating condition of an outboard blanket segment. Furthermore, the mechanical loads due to coolant pressure in normal operating conditions have been also taken into account. All the above loads have been combined as an input for a FEM stress analysis and the resulting stress distribution has been evaluated. Finally, the structure design and Pb-17Li flow velocity has been optimized according to the calculations and analysis.     

Development of the water cooled lithium lead blanket for DEMO

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

Neutronic analysis of an inertial fusion energy breeder with SiCf/SiC

Neutronic performance of a blanket driven ICF (Inertial confinement fusion) neutron based on SiC_f/SiC composite material is investigated for fissile fuel breeding. The investigated blanket is fueled with ThO₂ and cooled with natural lithium or (LiF)₂BeF₂ or Li₁₇Pb₈₃ or ⁴He coolant. MCNP4B Code is used for calculations of neutronic data per DT neutron. Calculations have show that values of TBR (tritium breeding ratio) being one of the main neutronic paremeters of fusion reactors are greater than 1.05 in all type of coolant, and the breeder hybrid reactor is self-sufficient in the tritium required for the DT fusion driver. Calculations show that natural lithium coolant blanket has the highest TBR (1.298) and M (fusion energy multiplication) (2.235), Li₁₇Pb₈₃ coolant blanket has the highest FFBR (fissile fuel breeding ratio) (0.3489) and NNM (net neutron multiplication) (1.6337). ⁴He coolant blanket has also the best Γ (peek-to-average fission power density ratio) (1.711). Values of neutron leakage out of the blanket in all type of coolants are quite low due to SiC reflector and B₄C shielding.     

Analysis and optimization of the manifolds of the HCCB blanket for CFETR

Thermal-hydraulic performance is a challenging issue in helium-cooled ceramic breeder (HCCB) blanket design due to the extremely complicated working environment and the strict limits of materials temperature. The heat loads deposited on the HCCB blanket comprises of severe surface heat flux from plasma and the volumetric nuclear heat from neutron irradiation, which can be exhausted by the built-in cooling channels of the components. High pressure helium with 8 MPa, distributed from the coolant manifolds is employed as coolant in the blanket. The design and optimization of the manifolds configuration was performed to guarantee the accurate flow control of helium coolant. The flow distribution in the coolant manifolds was investigated based on the structural improvement of manifolds aiming at overall uniform mass flow rates and better flow streamline distribution without obvious vortexes. The peak temperature of different functional materials in the blanket under normal operating condition is below allowable material limits. It is found that the components in the current blanket module could be cooled effectively under the intense thermal loads due to the updated design and optimization analysis of manifolds.     

Preliminary numerical analysis of MHD flow of DFLL-TBM on the single coolant stage

The liquid lithium–lead (PbLi) breeder blanket concept has been explored extensively due to their potential attractiveness. To check and validate the feasibility, the China dual-functional lithium lead test blanket module (DFLL-TBM) system, which is designated to demonstrate the integrated technologies of both He single coolant (SLL: single-cooled lithium lead) and He–LiPb dual-coolant (DLL: dual-cooled lithium lead) blankets, is proposed for test in ITER. One of the key feasibility issues is the impact of liquid metal MHD effect which will influence the pressure drop, flow distribution, and heat transfer in a DFLL-TBM.To reduce MHD effect, an electrically insulating coating is applied onto the inner surface of the flow channel for single coolant blanket. In this work, a preliminary numerical study of MHD flows in a simplified DFLL-TBM model on the single coolant stage has been carried out to assess the performance of such a concept with regard to the above mentioned MHD problems and constraints. The flow distribution and MHD pressure drop of LiPb flow in the SLL stage TBM are analyzed.     

Design approach of the ARIES-AT power core and vacuum vessel cost assessment

The complexity of fusion power plants require the integration of many diverse and important system requirements to achieve a design approach that is viewed as a commercially viable electric plant. The ARIES-AT power core design builds upon a history of fusion power core designs that evolve along with physics and engineering advances. The baseline design point is optimized for maximum performance and minimum capital cost based upon the ARIES systems code results, along with physics and engineering analyses. The ARIES-AT power core is designed to be quick and easily maintainable to achieve high plant availability. A key element to achieve the high availability is the integration of the core elements with the design of the vacuum vessel. The vacuum vessel design is developed in more detail to assure the key assembly and maintenance features could be realized at an affordable cost.