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Linkenheim-Hochstetten, Germany

Wang X.R.,University of California at San Diego | Tillack M.S.,University of California at San Diego | Koehly C.,Karlsruhe Institute of Technology | Malang S.,Fusion Nuclear Technology Consulting | And 2 more authors.
Fusion Science and Technology | Year: 2015

ARIES-ACT1 engineering design efforts were devoted to developing a credible configuration that allows for rapid removal of full-power core sectors followed by disassembly in hot cells during maintenance. The power core evolved with the main objective of achieving high performance while maintaining attractive design features and credible configuration, maintenance, and fabrication processes. To achieve high availability and maintainability of a fusion power plant, the power core components of a sector, including inboard and outboard first wall (FW)/blankets, upper and lower divertors, and structural ring or hightemperature shield, were integrated into one replacement unit to minimize time-consuming handling inside the plasma chamber. As with the ARIES-AT design, the FW/blanket design was based on Pb-17Li as coolant and breeder, and low-activation SiC/SiC as structural material; however, the Pb-17Li mass flow rate control, flow path, FW and blanket cooling channels, coolant access pipes, and blanket structural configuration have been revised and improved to provide about the same thermal performance (∼58% thermal efficiency) while keeping the magnetohydrodynamic pressure drop and pumping power, material temperature, and stresses at an acceptable level. Helium-cooled W or W-alloy divertor concepts were developed to accommodate a peak surface heat flux up to ∼14 MW/m2. They include a smaller finger-based divertor and a midsized T-tube and larger plate-type divertor concepts, which take advantage of a simple configuration, and the smaller number of plate units and joints in a power plant. The two-zone divertor concept, with the combination of a finger-based divertor and plate-type divertor, was selected and integrated into the ARIES-ACT1 power core. The fingers are used to accommodate the designed peak heat flux of ∼13 MW/m2, while the plate-type divertor is used for the lower heat flux region. The overall power core configuration and system integration, as well as the definitions of major power core components, such as the FW/blankets, divertor, structural ring, and the vacuum vessel, are described here and the main design features are highlighted. Sector maintenance operations have been investigated and motion demonstrations for removing the power core sectors have been performed using state-of-the-art three-dimensional CAD to analyze the clearances and spaces in all directions. The maintenance sequence and procedure for removing the replacement unit from the plasma chamber to the hot cell for exchange and refurbishment are also discussed in this paper. Source


Tillack M.S.,University of California at San Diego | Wang X.R.,University of California at San Diego | Malang S.,Fusion Nuclear Technology Consulting | Najmabadi F.,University of California at San Diego
Fusion Science and Technology | Year: 2013

ARIES-ACT1 is an advanced tokamak power plant conceptual design that utilizes SiC composite structural material in the blanket and PbLi as the tritium breeder and coolant. This design concept represents an evolutionary step from ARIES-AT, which has guided tokamak research programs for the past decade. In conjunction with a helium Brayton power cycle, the high primary coolant outlet temperature allows thermal conversion efficiency of 58%. The self-cooled blanket and He-cooled W-alloy divertor provide the ability to survive relatively high power density with acceptable projected lifetime. In ARIES-ACT1, we attempted to add "robustness" to the design point without major sacrifices in performance. In this paper, we will discuss the main features of the power core and selected details in the design and analysis. Source


Tillack M.S.,University of California at San Diego | Humrickhouse P.W.,Idaho National Laboratory | Malang S.,Fusion Nuclear Technology Consulting | Rowcliffe A.F.,Oak Ridge National Laboratory
Fusion Engineering and Design | Year: 2015

Water has both advantages and disadvantages as a coolant in conceptual designs of future fusion power plants. In the United States, water has not been chosen as a fusion power core coolant for decades. Researchers in other countries continue to adopt water in their designs, in some cases as the leading or sole candidate. In this article, we summarize the technical challenges resulting from the choice of water coolant and the differences in approach and assumptions that lead to different design decisions amongst researchers in this field. © 2014 Elsevier B.V. Source


El-Guebaly L.A.,University of Wisconsin - Madison | Jaber A.,University of Wisconsin - Madison | Malang S.,Fusion Nuclear Technology Consulting
Fusion Science and Technology | Year: 2012

There is a strong indication that the dual-cooled LiPb blanket is the preferred concept for many fusion power plants being designed around the world. The ability of the blanket to provide tritium self-sufficiency is among the important issues that we investigated in detail for ARIES-ACT to pinpoint the design elements that degrade the breeding the most, using state-of-the-art neutronics codes. A novel stepwise approach was developed to identify the exact cause of the degradation in the tritium breeding ratio (TBR), initially 1.8 for an ideal system, reaching 1.05 for a practical design. More broadly, this paper gives many insights into the impact that internal components of the blanket as well as essential parts of a tokamak can have on the TBR and the more damaging or enhancing conditions or changes to the breeding. To overcome the challenges of dealing with all tritium-related uncertainties in several subsystems, we suggest adjusting the Li enrichment online during operation to mitigate concerns about the danger of placing the plant at risk due to tritium shortage as well as the problem of handling and safeguarding any surplus of tritium. Source


El-Guebaly L.,University of Wisconsin - Madison | Huhn T.,University of Wisconsin - Madison | Rowcliffe A.,Oak Ridge National Laboratory | Malang S.,Fusion Nuclear Technology Consulting
Fusion Science and Technology | Year: 2013

Research has been conducted to find the optimal steel to use in the vacuum vessel (VV) of ARIES power plants. The VV should meet several design criteria, including activation and fabrication requirements. Seven different types of steel were examined in order to determine which steel would be the best candidate for the ARIES VV. The main concerns are related to activation, properties under irradiation, and fabrication of a sizable VV. Steels generating high-level waste (such as 316-SS) were excluded from possible material choices. As a VV material, there is the necessity for a carefully controlled the post-weld-heat-treatment at ∼750°C after assembly, welding, and rewelding. For this particular reason, the F82H FS is not suitable for the ARIES VV. The newly developed 3Cr-3WV bainitic FS meets the activation requirements and has the potential to satisfy the fabrication requirements for the ARIES VV. It is recommended for further consideration because of several advantages over other candidate steels. Source

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