E.P. Marriott

The Science and Technologies for Fusion Energy With Lasers and Direct-Drive Targets

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.

Neutronics performance parameters for the US dual coolant lead lithium ITER test blanket module

Neutronics analysis was performed for the reference design of the US dual coolant lead lithium (DCLL) ITER test blanket module (TBM). Detailed CAD models were utilized in the analysis. Relevant nuclear performance parameters were determined. These include tritium breeding, nuclear heating, radiation damage, transmutations, and shielding requirements. The calculated tritium breeding ratio (TBR) in the DCLL TBM is 0.561 and the total nuclear heating is 0.574 MW. For the ITER fluence goal of 0.3 MWa/m2, the peak cumulative radiation damage and He production in the first wall (FW) are 5.1 dpa and 56 appm, respectively. Mg builds up in the SiC flow channel inserts (FCI) to ∼100 appm with possible impact on electrical and thermal conductivities. About 1.2 m thick shield is required behind the TBM to allow personnel access for maintenance.

Application of CAD-neutronics coupling to geometrically complex fusion systems

An innovative computational tool (DAG-MCNP) has been developed for efficient and accurate 3-D nuclear analysis of geometrically complex fusion systems. Direct coupling with CAD models allows preserving the geometrical details, eliminating possible human error, and faster design iterations. DAG-MCNP has been applied to perform 3-D nuclear analysis for several fusion designs and demonstrated the ability to generate high-fidelity high-resolution results that significantly improve the design process. This tool will be the core for a full CAD-based simulation predictive capability that couples engineering analyses directly to the CAD solid model.

Progress on DCLL Blanket Concept

Abstract 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 (~700°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.

Neutronics Analysis in Support of the Fusion Development Facility Design Evolution

Abstract 3-D neutronics analysis was performed for the baseline design of FDF. Two blanket concepts were considered; Dual Coolant Lead Lithium (DCLL), and Helium Cooled Ceramic Breeder (HCCB). A peak outboard neutron wall loading of 2 MW/m2 and a fluence of 6 MW-yr/m2 can be achieved with 240 MW fusion power. The tritium breeding ratio is adequate for both blankets. Modest magnet damage parameters were obtained. However, it is recommended that the PF coil in the divertor region be moved vertically farther from the mid-plane to allow adding ~15 cm of shield to reduce the peak organic insulator dose to an acceptable level.

Laser IFE Direct Drive Chamber Concepts with Magnetic Intervention

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.

Mobile tiles for inertial fusion first wall/blanket

A conceptually simple first wall (FW) and blanket design for an inertial fusion system based on utilizing mobile FW tiles is presented. Using these tiles that are periodically removed, annealed, and reinstalled, tritium retention and surface erosion concerns for inertial fusion FW could be mitigated. A conceptual configuration has been developed with consideration for laser beam port accommodation and a simple tile insertion and removal scheme. Tritium self-sufficiency can be achieved with a variety of options. The current preferred design option utilizes liquid Li breeder that serves also as coolant for both the FW tiles and blanket with ferritic steel structural material. Thermal analysis for the carbon fiber composite FW tiles in the HAPL nuclear environment indicates that the maximum temperature will be ∼1300°C.

Nuclear Heating and Radiation Damage in the NBI Region of the ITER Vacuum Vessel

Abstract The ITER vacuum vessel (VV) is a double walled toroidal shaped stainless steel structure divided into nine 40 degree sectors. In the design process for the ITER blanket system (which provides shielding for the VV), determining integrated nuclear heating loads on the VV is important for cooling system sizing and determining localized nuclear heating on the VV is important for assessing thermal stress loads. Further, determining radiation damage, displacements per atom (dpa) on the VV, is important in meeting pressure vessel limits. Near the neutral beam injection (NBI) region of the VV (both sector 2 and sector 3), there are significant gaps and cut-outs in the blanket system to accommodate the 3 heating neutral beam (HNB) ports and the diagnostic neutral beam (DNB) port. These features lead to higher localized radiation loads. Previous analysis indicated high nuclear heating and dpa in the NBI region. The CAD based DAG-MCNP5 transport code was used to perform neutronics calculations in detailed, updated CAD models of the NBI region. For this work, a 40 degree model of sector 2 (which includes the HNB1 port, the DNB port, and the HNB2 port) was analyzed. Three design options were investigated which add shielding in the DNB port region by using port liners. Mesh tally maps of both nuclear heating and dpa are provided for the VV in the BM13-16 region. Peak dpa values ranged from 0.41–0.65 dpa. Two of the 3 design options investigated had peak dpa values near the DNB port within the ITER dpa limit of 0.5 dpa. Peak nuclear heating results ranged from 1.7 W/cm3 to 2.0 W/cm3. The mesh tally maps of nuclear heating have been provided to the ITER Organization for subsequent finite element engineering analysis. Preliminary analysis has shown the thermal stress levels are unacceptable with the added shielding. The results of this work are being used by the ITER Blanket and Tokamak Integration groups to assess the current design and modify blanket module (BM) design where needed if radiation loads are excessive.

Improved geometrical design of the US DCLL ITER test blanket module

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.

Nuclear heating in critical components of alternative ITER first wall attachment mechanism

A nuclear analysis was performed for an alternate ITER first wall attachment scheme to estimate the nuclear heating in critical components. With a variety of components near the first wall without direct cooling, there is concern that the operating temperatures will be above engineering limits. Using DAG-MCNP5, a 1-D analysis framework was used to estimate the nuclear heating in a detailed 3-D model of the first wall and shield attachment mechanism. The parts of the stainless steel hinge at the base of system experience nuclear heating between 1.85 and 5.07 W/cm3. The copper yoke and Inconel-718 yoke pin experience a heat rate of 2.44 W/cm3. If stainless steel bolts are used, the bolt heads will experience 2.9 W/cm3, with only a small reduction from plugging the bolt access holes. If molybdenum bolts are used, the bolt head heating is increased to 3.4 W/cm3. The heating in the hinge parts and stainless steel bolts is expected to lead to temperatures higher than engineering limits. Due to a higher thermal conductivity and melting temperature, the Mo bolts are expected to operate at acceptable temperatures.