Fusion Nuclear Technology Consulting

Linkenheim, Germany

Fusion Nuclear Technology Consulting

Linkenheim, Germany
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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.


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-ACT2 is a conventional tokamak power plant conceptual design that uses a dual-coolant lead-lithium (DCLL) blanket concept with a RAFS (reduced-activation ferritic steel) first-wall (FW) and blanket structure. The design concept is the first fully integrated study of the DCLL blanket in a tokamak power plant. The major engineering efforts were to develop a credible configuration that can meet aggressive maintenance goals and achieve high availability and maintainability; to design a DCLL blanket that can meet tritium breeding requirements with reasonable helium and Pb-17Li cooling schemes to remove the surface and volumetric thermal power in the blanket while keeping the helium pressure drop, magneto-hydrodynamic (MHD) pressure drop, and total pumping power low, and material temperatures and stresses at an acceptable level; to design manifolding and access pipes to connect/disconnect the inboard and outboard blanket sectors to the ring headers located underneath the reactor without affecting maintenance operations and creating major MHD effects when feeding all the Pb-17Li /He mass flow. Detailed three-dimensional finite element analysis of the DCLL blankets together with design iterations have been performed to finalize and optimize the major design parameters of the FW and blanket structure. The helium-cooled W plate-type divertor concept was adopted and integrated into the ACT2 DCLL power core to accommodate the peak surface heat flux of ∼10 MW/m2 predicted by edge plasma physics.


Tillack M.S.,University of California at San Diego | Raffray A.R.,University of California at San Diego | Raffray A.R.,ITER Organization | Wang X.R.,University of California at San Diego | And 4 more authors.
Fusion Engineering and Design | Year: 2011

Several advanced He-cooled W-alloy divertor concepts have been considered recently for power plant applications. They range in scale from a plate configuration with characteristic dimension of the order of 1 m, to the ARIES-CS T-tube configuration with characteristic dimension of the order of 10 cm, to the EU FZK finger concept with characteristic dimension of the order of 1.5 cm. The trend in moving to smaller-scale units is aimed at minimizing the thermal stress under a given heat load; however, this is done at the expense of increasing the number of units, with a corresponding impact on the reliability of the system. The possibility of optimizing the design by combining different configurations in an integrated design, based on the anticipated divertor heat flux profile, also has been proposed. Several heat transfer enhancement schemes have been considered in these designs, including slot jet, multi-hole jet, porous media and pin arrays. This paper summarizes recent US efforts in this area, including optimization and assessment of the different concepts under power plant conditions. Analytical and experimental studies of the concepts and cooling schemes are presented. Key issues are identified and discussed to help guide future R&D, including fabrication, joining, material behavior under the fusion environment and impact of design choice on reliability. © 2010 Elsevier B.V.


Wang X.R.,University of California at San Diego | Malang S.,Fusion Nuclear Technology Consulting | Tillack M.S.,University of California at San Diego | Burke J.,University of California at San Diego
Fusion Engineering and Design | Year: 2012

A number of advanced helium-cooled W-based divertor concepts have been proposed recently for fusion power plant applications within the framework of the ARIES Program. This paper summarizes design optimization and improvements of these concepts based on the minimum and maximum operating temperature of the W structure, pumping power and structural design limits. Re-evaluations of all concepts were performed with increased minimum operating temperature of the W structure from 700 °C to 800 °C in order to avoid embrittlement by neutron radiation. Design adjustments to allow for non-uniform heat flux profiles also have been considered. Comprehensive 3D thermal-fluid and 3D finite element thermo-mechanical analyses have been performed considering both elastic and plastic behavior and results are summarized in this paper. © 2012 Elsevier B.V. All rights reserved.


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.


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.


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.


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.


Wang X.R.,University of California at San Diego | Tillack M.S.,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 power plant has been designed and configured to allow for rapid removal of full power core sectors followed by disassembly in hot cells during maintenance operation. To achieve high availability and maintainablity of a fusion power plant, power core components of a sector, including inboard and outboard FW/blankets, upper and lower divertor, structural ring or high temperature shield were integrated into one replacement unit to minimize time comsuming handling inside plasma chamber. In this paper, the overall power core configuration and system integration, as well as the definitions of major power core components are described and main design features are highlighted.


Navaei D.,University of California at San Diego | Wang X.R.,University of California at San Diego | Tillack M.S.,University of California at San Diego | Malang S.,Fusion Nuclear Technology Consulting
Fusion Engineering and Design | Year: 2013

The use of tungsten as a plasma-facing material necessitates a transition joint to the oxide dispersion strengthened (ODS) steel or ferritic steel (FS) structural material of the primary coolant loop at the end of the divertor target plate where the surface heat flux is very low. A critical issue in the transition joints is the coefficient of thermal expansion (CTE) mismatch between the tungsten (or tungsten-alloy) and ODS steel, which can lead to unacceptably high thermal stresses during steady state and ratcheting during cyclic loads. Detailed 2D and 3D thermo-mechanical analyses were conducted to study the behavior of a transition from tungsten to FS with an intermediate layer of tantalum, located outside of the high heat flux region. The results include plastic strains under various loading conditions including fabrication processes, warm and cold shutdown, and allow for plastic behaviors leading to stress relaxation. The accumulation of plastic deformation may cause ratcheting. Modifications were proposed to the transition joint design in order to eliminate stress concentration and ratcheting under cyclic loading. The results of the modified design exhibited less plastic deformation in the joints as well as no ratcheting caused by warm and cold shutdown. © 2013 Elsevier B.V.

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