Neutronics analysis was performed on the latest Fusion Energy System Studies-Fusion Nuclear Science Facility (FESS-FNSF) design, which determined the neutron wall loading, tritium breeding ratio, and radiation damage. Sixteen different sectors configurations were investigated, with the main focus on determining the impact which each has upon the tritium breeding ratio (TBR) of the whole facility. This paper describes the stages of the nuclear analysis that serve to prove the radiation derived attributes of the system. (C) 2017 The Authors. Published by Elsevier B.V.

Two issues related to neutronics analysis of fusion systems were addressed for the purpose of physical design iterations as well as plant operation: tritium self-Psufficiency and shielding of the inboard magnet. State-Pof-Pthe-Part modeling/analysis tools facilitated a full 3-PD neutronics analysis of the latest FESS-PFNSF design. The first stage of the analysis involved the selection of materials for the first wall and blanket along with shielding materials to protect the magnet based on extensive 1-PD analyses. The second stage is a stepwise workflow to estimate the overall tritium breeding ratio with high fidelity. It involved a bottom-Pup approach by coupling the CAD model with the 3-PD MCNP code using DAGMC and adding the relevant design details in steps to assess the effect of such details on the tritium breeding ratio. The final stage involved calculations of the values of damage parameters at specific components: the first wall, the vacuum vessel, and magnet.

The Fusion Nuclear Science Facility (FNSF) is an essential element of the U.S. developmental roadmap to fusion energy. The facility displays the complex integration of tokamak components and subsystems in fusion environment while testing and developing fusion technologies for the U.S. demonstration (DEMO) plant. The integration of the neutronics, shielding, and activation assessments is a key element to the success of FNSF operation. This paper overviews the engineering aspects of the tokamak-based FNSF study and presents an integral scheme that considered the overall configuration, radiation limits, top-level design requirements (including maximizing the tritium breeding ratio), smart selection of low-activation materials for each component, radial build optimization and definition, environmental and safety constraints, and upper temperatures for the reuse of reduced activation ferritic martensitic and bainitic structures after severe loss of coolant accidents.

A fusion nuclear science facility (FNSF) could play an important role in the development of fusion energy by providing the nuclear environment needed to develop fusion materials and components. The spherical torus/tokamak (ST) is a leading candidate for an FNSF due to its potentially high neutron wall loading and modular configuration. A key consideration for the choice of FNSF configuration is the range of achievable missions as a function of device size. Possible missions include: providing high neutron wall loading and fluence, demonstrating tritium self-sufficiency, and demonstrating electrical self-sufficiency. All of these missions must also be compatible with a viable divertor, first-wall, and blanket solution. ST-FNSF configurations have been developed simultaneously incorporating for the first time: (1) a blanket system capable of tritium breeding ratio TBR approximate to 1, (2) a poloidal field coil set supporting high elongation and triangularity for a range of internal inductance and normalized beta values consistent with NSTX/NSTX-U previous/planned operation, (3) a long-legged divertor analogous to the MAST-U divertor which substantially reduces projected peak divertor heat-flux and has all outboard poloidal field coils outside the vacuum chamber and superconducting to reduce power consumption, and (4) a vertical maintenance scheme in which blanket structures and the centerstack can be removed independently. Progress in these ST-FNSF missions versus configuration studies including dependence on plasma major radius R-0 for a range 1 m-2.2 m are described. In particular, it is found the threshold major radius for TBR = 1 is R-0 >= 1.7 m, and a smaller R-0 = 1 m ST device has TBR approximate to 0.9 which is below unity but substantially reduces T consumption relative to not breeding. Calculations of neutral beam heating and current drive for non-inductive ramp-up and sustainment are described. An A = 2, R-0 = 3 m device incorporating high-temperature superconductor toroidal field coil magnets capable of high neutron fluence and both tritium and electrical self-sufficiency is also presented following systematic aspect ratio studies.

Under the US Fusion Nuclear Science and Technology Development program, we have selected the Dual Coolant Lead Lithium concept (DCLL) as a reference blanket, which 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. The self-cooled breeder PbLi is circulated for power conversion and for tritium breeding. A SiC-based flow channel insert (FCI) is used as a means for magnetohydrodynamic pressure drop reduction from the circulating liquid PbLi and as a thermal insulator to separate the high-temperature PbLi (similar to 700 degrees C) from the helium-cooled RAF/M steel structure. We are making progress on related R&D needs to address critical Fusion Nuclear Science and Facility (FNSF) and DEMO blanket development issues. While performing the function as the Interface Coordinator for the DCLL blanket concept, we were developing the mechanical design and performing neutronics, structural and thermal hydraulics analyses of the DCLL TBM module. We estimated the necessary ancillary equipment that will be needed at the ITER site, and a detailed safety impact report was prepared. This provided additional understanding of the DCLL blanket concept in preparation for the FNSF and DEMO. This paper is a summary report on the progress of the DCLL TBM design and R&D for the DCLL blanket concept.

Computational fluid dynamics simulations have demonstrated flow problems within the helium flow path in the current US DCLL ITER test blanket module design. New geometry for the helium flow path has been designed that will improve flow evenness and simplify the overall helium flow path within the test blanket module while maintaining the overall test blanket module geometry. Global changes to the test blanket module geometry can be implemented based on these improvements.

We are carrying out a multidisciplinary multi-institutional program to develop the scientific and technical basis for inertial fusion energy (IFE) based on laser drivers and direct-drive targets. The key components are developed as an integrated system, linking the science, technology, and final application of a 1000-MWe pure-fusion power plant. The science and technologies developed here are flexible enough to be applied to other size systems. The scientific justification for this work is a family of target designs (simulations) that show that direct drive has the potential to provide the high gains needed for a pure-fusion power plant. Two competing lasers are under development: the diode-pumped solid-state laser (DPPSL) and the electron-beam-pumped krypton fluoride (KrF) gas laser. This paper will present the current state of the art in the target designs and lasers, as well as the other IFE technologies required for energy, including final optics (grazing incidence and dielectrics), chambers, and target fabrication, injection, and tracking technologies. All of these are applicable to both laser systems and to other laser IFE-based concepts. However, in some of the higher performance target designs, the DPPSL will require more energy to reach the same yield as with the KrF laser.

The High Average Power Laser (HAPL) program is focusing on the development of laser IFE power plants based on lasers, direct-drive targets and dry wall chambers. One key issue is the survival of the chamber wall under the ion threat spectra (representing ˜25% of the yield energy). The possibility of steering the ions away from the chamber to specially-designed dump chambers using magnetic intervention is being investigated. This brings up the intriguing possibility of utilizing a liquid wall to accommodate the ion fluxes in the dump chamber provided the right measures are taken to prevent the liquid from contaminating the main chamber. This paper covers the initial assessment of different magnetic configurations for a laser IFE chamber. Their key characteristics are described; results of the supporting design analyses are summarized; and the major findings and issues are highlighted.