In the ST reactor, the radial build of TF coil and the shield play a key role in determining the size of a reactor. For self-consistent determination of these components and physics parameters, a system analysis code is coupled with the onedimensional radiation transport code. Conceptual design of a compact superconducting ST reactor with aspect ratio of less than 2 was conducted and it is shown that the ST reactor with outboard blanket only can provide tritium selfsufficiency by using an inboard neutron reflector instead of breeding blanket. With the use of an improved shielding material and high temperature superconducting magnets with high critical current density open up the possibility of a fusion power plant with compact size and smaller auxiliary heating power simultaneously at low aspect ratio.

An optimum plasma pressure/current density profile and corresponding heating/current drive (H/CD) determination scheme is newly developed by integrating equilibrium, stability, confinement, and H/CD, self-consistently subject to maximize the fusion gain for Korean fusion demonstration reactor (K-DEMO) steady-state operation scenarios. The integrated plasma modeling package, FASTRAN/IPS, is adopted for the integrated numerical apparatus. The target pressure profile with a pedestal structure is investigated by varying its peaking, pedestal height and width as a first step. Formation of stable equilibria is evaluated by solving the Grad–Shafranov equation and checking linear MHD stability. For the case of potentially stable equilibrium, required external heating distribution is calculated by considering both power balance and external current drive alignment to reproduce the pressure profile of the stable equilibrium. Electron/ion temperature and poloidal flux evolutions are solved with the derived heating configuration to find a steady-state scenario and achieve self-consistent plasma profiles. A self-consistent target steady-state pressure and current profile parameters are proposed through designed systematic algorithm with fusion power PF  =  2070 MW, fusion gain Q  =  19.7, and normalized beta βN  =  2.8 at toroidal field BT  =  7.4 T and plasma current IP  =  15.5 MA. Feasibility of fusion power PF  =  3000 MW operation is also explored with enhanced density and temperature limit assumption.

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This paper describes benchmark calculations for the APR1400 nuclear reactor performed using the high-fidelity deterministic whole-core simulator MPACT compared to reference solutions generated by the Monte Carlo code McCARD. The methodology presented in this paper is a common approach in the field of nuclear reactor analysis, when measured data are not available for comparison, and may be more broadly applied in other simulation applications of energy systems. The benchmark consists of several problems that span the complexity of single pins to a hot full power cycle depletion. Overall, MPACT shows excellent agreement compared to the reference solutions. MPACT effectively predicts the reactivity for different geometries and several temperature and boron conditions. The largest deviation from McCARD occurs for cold zero conditions in which the fuel, moderator, and cladding are all 300 K. Possible reasons for this are discussed. Excluding these cases, the ρ reactivity difference from McCARD is consistently below 100 pcm. For single fuel pin problems, the highest error of 151 pcm occurs for the lowest fuel enrichment of 1.71 wt.% UO2, indicating possible, albeit small, enrichment bias in MPACT’s cross-section library. Furthermore, MOC and spatial mesh parametric studies indicate that default meshing parameters and options yield results comparable to finely meshed cases. Additionally, there is very good agreement of the radial and axial power distributions. RMS radial pin and assembly power differences for all cases are at or below 0.75%, and all RMS axial power differences are below 1.65%. These results are comparable to previous results from the VERA progression problems benchmark and meet generally accepted accuracy criteria for whole-core transport codes.

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