Moving beyond ITER toward a compact magnetic fusion demonstration reactor (Demo) will require the integration of high plasma performance in steady-state with advanced methods for dissipating very high divertor heat-fluxes, while respecting strict limits on tritium retention. Expressing power exhaust requirements in terms of P_heat/R, future ARIES reactors are projected to operate with 60-200MW/m, Component Test Facilities (CTF) 40-50MW/m, and ITER 20-25MW/m. However, new and planned long-pulse experiments (such as EAST, JT60-SA, KSTAR, SST-1) are currently projected to operate at values of up to 16MW/m. The considerable gap between upcoming experiments and a CTF or fusion power plant motivates the proposal of a new experiment – the National High-power advanced-Torus eXperiment (NHTX) – whose mission is to study the integration of high-confinement, high-beta, long-pulse fully-non-inductive plasma operation with a fusion-relevant high-power plasma-boundary interface.  Systems code studies find an optimal aspect ratio A=1.8-2 simultaneously maximizes the achievable P/R and non-inductive I_P (bootstrap + neutral beam current drive). The PPPL site power and TFTR test cell and neutral beams are well suited to the NHTX mission, and with P_AUX = 50MW and R₀=1m achieves P/R = 50MW/m.  The resultant initial NHTX design point is A=1.8, R₀=1m, I_P=3-4MA, B_T=2T, κ=2.7-3, HH_98Y = 1.3, β_N=4.5, β_T=15%, f_GW=0.4-0.5, f_BS≥ 65%, f_NI = 100%, τ_pulse up to 1000s, and Τ_wall ~600 °C for hydrogenic isotope retention studies using a range of plasma facing materials, including liquid metals.  A highly flexible divertor coil set is a crucial design element which facilitates testing of many divertor geometries including an ITER-like divertor and a wide range of poloidal flux expansion = 3-30.  TRANSP simulations of the beam-driven current, the role of other possible current-drive sources, and future engineering and physics analysis work will be discussed.