There is no practical external source of tritium for fusion energy development beyond ITER and all subsequent fusion systems have to breed their own tritium. To ensure tritium self-sufficiency, the ...calculated achievable tritium breeding ratio (TBR) should be equal to or greater than the required TBR. The potential of achieving tritium self-sufficiency depends on many system physics and technology parameters. Interactive physics and technology R&D programs should be implemented to determine the potential of realizing those physics and technology options and parameters that have large effects on attaining a realistic “window” for tritium self-sufficiency. The ranges of plasma and technology conditions that need to be met, in order to ensure tritium self-sufficiency, are identified.
The amount and type of metallic transmutants produced in SiC/SiC when used in magnetic (MFE) and inertial (IFE) confinement fusion systems are determined and compared to those obtained following ...irradiation in fission reactors. Up to ∼1.3% metallic transmutants are generated at the expected lifetime of the fusion blanket. Irradiation in fission reactors to the same fast neutron fluence produces about an order of magnitude lower metallic transmutation products than in fusion systems. While the dominant component in fusion systems is Mg, P is the main transmutation product in fission reactors. The impact on the SiC/SiC properties is not fully understood. The results of this work will help guide irradiation experiments in fission reactors to properly simulate the conditions in fusion systems by possible ion implantation. In addition, the results represent a necessary input for modeling activities aimed at understanding the expected effects on properties.
The amount and type of gaseous and metallic transmutants produced in tungsten (W) when used as a plasma-facing armor in magnetic (MFE) and inertial (IFE) confinement fusion systems were determined ...and compared to those obtained following irradiation in fission reactors. Up to ∽8% metallic transmutants are generated at the expected lifetime of the fusion blanket. Irradiation in fission reactors to the same fast neutron fluence yields a much larger amount of metallic transmutation products than in fusion systems. While the dominant component in fusion systems is rhenium (Re), osmium (Os) is the main transmutation product in fission reactors. The impact on the W properties needs to be assessed. The results of this work will help guide irradiation experiments in fission reactors to properly simulate the conditions in fusion systems by possible direct implantation of transmutation products in irradiated samples. In addition, the results represent a necessary input for modeling activities aimed at understanding the expected effects on properties.
Radiation damage parameters in SiC/SiC composite structures are determined in both magnetic (MFE) and inertial (IFE) confinement fusion systems. Variations in the geometry, neutron energy spectrum, ...and pulsed nature of neutron production result in significant differences in damage parameters between the two systems. With the same neutron wall loading, the displacement damage rate in the first wall in an IFE system is ∼10% lower than in an MFE system, while gas production and burnup rates are a factor of 2 lower. Self-cooled LiPb and Flibe blankets were analyzed. While using LiPb results in higher displacement damage, Flibe yields higher gas production and burnup rates. The effects of displacement damage and helium production on defect accumulation in SiC/SiC composites are also discussed.
•A FNSF is needed to reduce the knowledge gaps to a fusion DEMO and accelerate progress toward fusion energy.•FNSF will test and qualify first-wall/blanket components and materials in a DEMO-relevant ...fusion environment.•The Advanced Tokamak approach enables reduced size and risks, and is on a direct path to an attractive target power plant.•Near term research focus on specific tasks can enable starting FNSF construction within the next ten years.
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=2.7m/0.77m, κ=2.3, BT=5.4T, IP=6.6 MA, βN=2.75, Pfus=127MW. The modest bootstrap fraction of ƒBS=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.
An overview of the US DCLL ITER-TBM program Wong, C.P.C.; Abdou, M.; Dagher, M. ...
Fusion engineering and design,
12/2010, Letnik:
85, Številka:
7
Journal Article, Conference Proceeding
Recenzirano
Under the US Fusion Nuclear Science and Technology program, we selected the Dual Coolant Lead Lithium (DCLL) concept as our primary Test Blanket Module (TBM) for testing in ITER. The DCLL blanket ...concept has the potential to be a high-performance DEMO blanket design with a projected thermal efficiency of >40%. Reduced activation ferritic/martensitic (RAF/M) steel is used as the structural material. Helium is used to cool the first wall and blanket structure, and the self-cooled Pb-17Li breeder is circulated for power conversion and for tritium extraction. A SiC-based flow channel insert (FCI) is used as an electrical insulator for magnetohydrodynamic pressure drop reduction from the circulating Pb-17Li and as a thermal insulator to separate the high-temperature Pb-17Li (∼650–700
°C) from the RAF/M structure, which has a corrosion temperature limit of ∼480
°C. The RAF/M material must also operate at temperatures above 350
°C but less than 550
°C. We are continuing the development of the mechanical design and performing neutronics, structural and thermal hydraulics analyses of the DCLL TBM module. Prototypical FCI structures were fabricated and further attention was paid to MHD effects and the design of the inboard blanket for DEMO. We are also making progress on related R&D needs to address key areas. This paper is a summary report on the progress and results of recent DCLL TBM development activities.
In ITER, the blanket modules (BM) are arranged around the plasma to provide thermal and nuclear shielding for the vacuum vessel (VV), magnets, and other components. As a part of the BM design ...process, nuclear analysis is required to determine the level of nuclear heating, helium production, and radiation damage in the BM. Additionally, nuclear heating in the VV is also important for assessing the BM design. We used the CAD based DAG-MCNP5 transport code to analyze detailed models inserted into a 40-degree partially homogenized ITER global model. The regions analyzed include BM01, the neutral beam injection (NB) region, and the upper port region. For BM01, the results show that He production meets the limit necessary for re-welding, and the VV heating behind BM01 is acceptable. For the NBI region, the VV nuclear heating behind the NB region exceeds the design limit by a factor of two. For the upper port region, the nuclear heating of the VV exceeds the design limit by up to 20%. The results presented in this work are being used to modify the BM design in the cases where limits are exceeded.
In this letter, we addressed the problem of estimating the time delay and the frequencies of noisy sinusoidal signals received at two spatially separated sensors. We employ the Propagator Method (PM) ...in conjunction with the well-known MUSIC/root-MUSIC algorithm; the proposed method would generate estimates of the unknown parameters. Such estimates are based on the observation and/or covariance matrices. Moreover, the PM does not require the eigenvalue decomposition (EVD) or singular value decomposition (SVD) of the cross-spectral matrix (CSM) of received signals; therefore, a significant improvement in computational load is achieved. Computer simulations are also included to demonstrate the effectiveness of the proposed method.
Within the magnetic fusion energy program in the US, a program called APEX is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around the plasma. As part ...of this work, the APEX Team has investigated several possible design implementations and developed a specific engineering concept for a fusion reactor with liquid walls. Our approach has been to utilize an already established design for a future fusion reactor, the ARIES-RS, for the basic chamber geometry and magnetic configuration, and to replace the chamber technology in this design with liquid wall technology for a first wall and divertor and a blanket with adequate tritium breeding. This paper gives an overview of one design with a molten salt (a mixture of lithium, beryllium and sodium fluorides) forming the liquid surfaces and a ferritic steel for the structural material of the blanket. The design point is a reactor with 3840
MW of fusion power of which 767
MW is in the form of energetic particles (alpha power) and 3073
MW is in the form of neutrons. The alpha plus auxiliary power total 909
MW of which 430
MW is radiated from the core mostly onto the first wall and the balance flows into the edge plasma and is distributed between the first wall and the divertor. In pursuing the application of liquid surfaces in APEX, the team has developed analytical tools that are significant achievements themselves and also pursued experiments on flowing liquids. This work is covered elsewhere, but the paper will also note several such areas to indicate the supporting science behind the design presented. Significant new work in modeling the plasma edge to understand the interaction of the plasma with the liquid walls is one example. Another is the incorporation of magneto-hydrodynamic (MHD) effects in fluid modeling and heat transfer.
Detailed 3-D neutronics calculations have been performed for the US DCLL TBM. The neutronics calculations were performed directly in the CAD model using the DAG-MCNP code that allows preserving the ...geometrical details. Detailed high-resolution, high-fidelity profiles of the nuclear parameters were generated using fine mesh tallies. These included tritium production, nuclear heating, and radiation damage. The TBM heterogeneity, exact source profile, and inclusion of the surrounding frame and other in-vessel components result in lower TBM nuclear parameters compared to the previous 1-D predictions. This work clearly demonstrates the importance of preserving geometrical details in nuclear analyses of geometrically complex components in fusion systems.
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