Publications

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In preparation for testing a lithium-helium heat exchanger at Sandia, unexpected rapid failure of the mild steel lithium preheater due to liquid metal embrittlement occurred when lithium at ~400 °C flowed into the preheater then at ~200 °C. This happened before the helium system was pressurized or heating with electron beams began. The paper presents an analysis of the preheater plus a discussion of some implications for fusion.

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The authors exposed a radiatively cooled, Li-filled tantalum (Ta) heat pipe (HP) to a H plasma in Magnum PSI continuously for ˜2 h. We kept the overall heat load on the inclined HP constant and varied the tilt to give peak heat fluxes of ˜7.5–13 MW/m2. The peak temperature reached ˜1250 °C. This paper describes the post-test analysis and discusses Li HPs with materials other than Ta for fusion. A companion paper describes the experiment.

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This review of the potential of robust plasma facing components (PFCs) with liquid surfaces for applications in future D/T fusion device summarizes the critical issues for liquid surfaces and research being done worldwide in confinement facilities, and supporting R&D in plasma surface interactions. In the paper are a set of questions and related criteria by which we will judge the progress and readiness of liquid surface PFCs. Part-II (separate paper) will cover R&D on the technology-oriented aspects of liquid surfaces including the liquid surfaces as integrated first walls in tritium breeding blankets, tritium retention and recovery, and safety.

A major challenge facing the design and operation of next-step high-power steady-state fusion devices is to develop a viable divertor solution with order-of-magnitude increases in power handling capability relative to present experience, while having acceptable divertor target plate erosion and being compatible with maintaining good core plasma confinement. A new initiative has been launched on DIII-D to develop the scientific basis for design, installation, and operation of an advanced divertor to evaluate boundary plasma solutions applicable to next step fusion experiments beyond ITER. Developing the scientific basis for fusion reactor divertor solutions must necessarily follow three lines of research, which we plan to pursue in DIII-D: (1) Advance scientific understanding and predictive capability through development and comparison between state-of-the art computational models and enhanced measurements using targeted parametric scans; (2) Develop and validate key divertor design concepts and codes through innovative variations in physical structure and magnetic geometry; (3) Assess candidate materials, determining the implications for core plasma operation and control, and develop mitigation techniques for any deleterious effects, incorporating development of plasma-material interaction models. These efforts will lead to design, installation, and evaluation of an advanced divertor for DIII-D to enable highly dissipative divertor operation at core density (n e/n GW), neutral fueling and impurity influx most compatible with high performance plasma scenarios and reactor relevant plasma facing components (PFCs). This paper highlights the current progress and near-term strategies of boundary/PMI research on DIII-D.