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As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy. A major design philosophy for the initiated conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) is engineering feasibility. A two-staged development plan is envisaged. K-DEMO is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used, in its initial stage, as a component test facility. Then, in its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electricity generation on the order of 500 MWe. After a thorough 0-D system analysis, the major radius and minor radius are chosen to be 6.8 m and 2.1 m, respectively. In order to minimize wave deflection, a top-launch high frequency (>200 GHz) electron cyclotron current drive (ECCD) system will be the key system for the current profile control. For matching the high frequency ECCD, a high toroidal field (TF) is required and can be achieved by using high current density Nb 3 Sn superconducting conductor. The peak magnetic field reaches to 16 T with the magnetic field at the plasma center above 7 T. Key features of the K-DEMO magnet system include the use of two TF coil winding packs, each of a different conductor design, to reduce the construction cost and save the space for the magnet structure material. [ABSTRACT FROM AUTHOR]

Abstract: A conceptual design study for a steady-state Korean fusion DEMO reactor (K-DEMO) has been initiated. Two peculiar features need to be noted. First, the major radius is designed to be just below 6.5m, considering practical engineering feasibilities. But still, high magnetic field at the plasma center around 8T is expected to be achieved by using current state-of-the-art high performance Nb3Sn strand technology. Second, a two-stage development plan is being considered. In the first stage, K-DEMO will demonstrate a net electricity generation but will also act as a component test facility. Then, after a major upgrade, K-DEMO is expected to show a net electric generation on the order of 300MWe and the competitiveness in cost of electricity (COE). Feasibility of such a practical, near-future demonstration reactor is studied in this paper, based on a zero dimensional system analysis code study. It was shown that a net electric generation on the order of 300MWe can be achieved below the optimistic β N limit of 5. The elongation of K-DEMO is around 1.8 with single null configuration. Detailed optimization process and the resultant various plasma parameters are described. [Copyright &y& Elsevier]

Abstract: As the ITER is being constructed, there is a growing anticipation for an earlier realization of fusion energy, so called fast-track approach. Korean strategy for fusion energy can be regarded as a fast-track approach and one special concept discussed in this paper is a two-stage development plan. At first, a steady-state Korean DEMO Reactor (K-DEMO) is designed not only to demonstrate a net electricity generation and a self-sustained tritium cycle, but also to be used as a component test facility. Then, at its second stage, a major upgrade is carried out by replacing in-vessel components in order to show a net electric generation on the order of 300MWe and the competitiveness in cost of electricity (COE). The major radius is designed to be just below 6.5m, considering practical engineering feasibilities. By using high performance Nb3Sn-based superconducting cable currently available, high magnetic field at the plasma center above 8 T can be achieved. A design concept for TF magnets and radial builds for the K-DEMO considering a vertical maintenance scheme, are presented together with preliminary design parameters. [Copyright &y& Elsevier]

Highlights: [•] One of blanket concepts under consideration in the K-DEMO is a water-cooled ceramic blanket. [•] The dependence of TBR on volume ratios of water and breeder has been investigated. [•] The possibility of use of D2O replacing H2O as a coolant has been investigated. [•] A comparative study through neutronics assessment has been performed with the MCNPX code. [Copyright &y& Elsevier]

A 16 Tesla, large bore superconducting magnet for testing the superconducting Cable-in-Conduit Conductor (CICC) is being designed in parallel with the design of the steady state Korea fusion demonstration reactor (K-DEMO) magnet system. The test facility magnet, named SUperConducting Conductor Experiment (SUCCEX), was designed in 2014 and some design modifications of the magnet have been conducted since 2020. The peak magnetic field of the K-DEMO toroidal field coil is about 16 T, and thus the required background field of the conductor test magnet is expected to be over 15 T in a large bore of 600 mm diameter. The background field will be in addition to the K-DEMO conductor sample's self-field. To reach the target magnetic field, high current density Nb3Sn strands (Jc > 1100 A/mm2 at 4.2 K, 16 T) were employed in the design of the high field CICC. The modified SUCCEX magnet consists of concentric solenoids, with a high field inner coil (IC) and low field outer coil (OC). They are connected in series with each other, and the operating current is about 24 kA. In this study, the overall design concept is presented in relation to the results of electro-magnetic, structural analyses. [ABSTRACT FROM AUTHOR]

A conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) was initiated in 2012. The superconducting magnet system is one of the key components of the K-DEMO and the preliminary study on superconducting magnet system has been done in 2015. The superconducting magnet system of K-DEMO consists of 16 TF (Toroidal Field), 8 CS (Central Solenoid) and 12 PF (Poloidal Field) coils. All of the TF, CS and PF coils use internally-cooled Cable-In-Conduit Conductors (CICC). A high performance Nb3Sn-based conductor, an ITER-type Nb3Sn and NbTi conductor will be used for the magnet. Based on the preliminary design, continuous design study is being performed and the K-DEMO magnet system has been evolved mainly for solving engineering issues. Two issues were considered for the TF, the reinforcement of the low field TF CICC structure and the fabrication stability of the high field TF CICC. In the CS magnet, reducing the number of Helium feed-throughs was a main issue. A small change has been made for the PF magnets for keeping a void fraction of the PF conductor. In this work, the major parameters and design updates of the K-DEMO magnet system is presented. [ABSTRACT FROM AUTHOR]

The Korean fusion demonstration reactor (K-DEMO) has completed a two year study looking at key Tokamak components and configuration options in preparation of a conceptual design phase. A key part of a device configuration centers on defining an arrangement that enhances the ability to reach high availability values by defining design solutions that foster simplified maintenance operations. To maximize the size and minimize the number of in-vessel components enlarged TF coils were defined that incorporate a pair of windings within each coil to mitigate pressure drop issues and to reduce the cost of the coils. A semi-permanent shield structure was defined to develop labyrinth interfaces between double-null plasma contoured shield modules, provide an entity to align blanket components and provide support against disruption loads—with a load path that equilibrates blanket, TF and PF loads through a base structure. Blanket piping services and auxiliary systems that interface with in-vessel components have played a major role in defining the overall device arrangement—concept details will be presented along with general arrangement features and preliminary results obtained from disruption analysis. [ABSTRACT FROM AUTHOR]

A pre-conceptual design study has been carried out for the Korean fusion demonstration reactor (K-DEMO) tokamak featured by high magnetic field ( B T0 = 7.4 T), R = 6.8 m, a = 2.1 m, and a steady-state operation. The design concepts of the K-DEMO blanket system considering the cooling in-vessel components with pressurized water and a solid pebble breeder are described herein. The structure of the K-DEMO blanket is toroidally subdivided into 16 inboard and 32 outboard sectors, in order to allow the vertical maintenance. Each blanket module is composed of plasma-facing first wall, layers of breeding parts, shielding and manifolds. A ceramic breeder using Li 4 SiO 4 pebbles with Be 12 Ti as neuron multiplier is employed for study. MCNP neutronic simulations and thermo-hydraulic analyses are interactively performed in order to satisfy two key aspects: achieving a global Tritium Breeding Ratio (TBR) >1.05 and operating within the maximum allowable temperature ranges of materials. [ABSTRACT FROM AUTHOR]

• A water-cooled tungsten monoblock divertor concept has been considered in the K-DEMO. • The detached divertor study has been carried out by the impurity seeding for N, Ne, and Ar to reduce the peak heat flux and particle flux. • The optimum designs were derived by the response surfaces optimization method for CuCrZr and RAFM steel, the major candidates for the heat sink in the K-DEMO. • The alternative divertor concepts were studied for the super X-divertor and snowflake divertor. The preliminary conceptual design of the Korean fusion demonstration reactor (K-DEMO) with a major radius of 6.8 m and the fusion power of 2200 MW has been studied since 2012. The overall configuration of the K-DEMO divetor system based on the ITER-like water-cooled tungsten technology is a double-null type symmetric divertor subdivided into 32 toroidal modules for the vertical maintenance. A detached divertor scenario with impurity seeding was considered as the primary approach for the power exhaust to reduce the peak heat flux lower than the engineering limit of 10 MW/m2. The power exhaust performance at the scrap off layer was estimated by using UEDGE-2D code, a two-dimensional fluid transport code for collisional edge plasma and neutral species like N, Ne, and Ar. Particle and heat flux on inboard and outboard divertor targets were calculated for the detached cased depending on parameters such as the impurity seeding rate, pumping rate, and the pedestal density. On the other hand, a magnetic solution like X-divertor, snowflake divertor, and super X-divertor to expand the plasma wet area was considered for K-DEMO since the detached divertor increasing a radiation fraction by impurity seeding might be able to be unstable. However, the extremely high current of poloidal coils was required more than the engineering limit, 20 MA, to form magnetic field lines for the alternative divertors. Based on the physical calculation of the edge plasma, engineering analyses were carried out to find out the thermal and structural reliability. The thermo-hydraulic analysis confirmed thermal stability, whether all comprising materials are operating within their allowable temperature windows when the case of the peak heat flux is set to 10 MW/m2 on the outboard divertor target. The response surface optimization method derived two optimal design candidates employing two kinds of heat sink materials, respectively: the reduced activation ferritic martensitic (RAFM) steel and CuCrZr alloy. The drawbacks and merits of the two materials were definite. The optimal design with applying RAFM steel was vulnerable to withstand thermal and mechanical loads since low thermal conductivity caused too thin thickness of the heat sink. On the other hand, the CuCrZr alloy has critical drawbacks in terms of activation and radioactive waste despite its high thermal conductivity. Meanwhile, preliminary electromagnetic (EM) analysis was carried out to estimate the EM loads caused by the abnormal behaviors of plasma since EM loads are one of the most critical external loads for designing a DEMO divertor. [ABSTRACT FROM AUTHOR]

• The design updated of the K-DEMO CS coils to minimize the pressure drop in a long cooling path. • A CFX simulation carried out to check the pressure drop condition using the ITER CS coil test results. • The allowable flux swing condition (dB/dt) investigated using the thermo-hydraulic analysis. • The simulation results satisfied the requirement of the flux swing (3.5 T/s) around the discharge time constant 3 s and the allowable dI/dt was 12.19 kA/s. The K-DEMO superconducting tokamak is being studied for the realization of the fusion plant since 2012. The preliminary design of the conductor of each magnet introduced in 2015 and the design are being updated for solving engineering issues. In the previous study, the K-DEMO CS(Central Solenoid) conductor was modified to reduce the number of the cooling channel. A central cooling spiral was applied to the CS conductor to minimize the pressure drop in a long cooling path and the shape of conductor was changed to square. This design change should guarantee to provide sufficient cooling conditions to reach the full flattop current under a wide range of conditions especially, the selected for the plasma ramp-up sequence development of CS coils. In this study, the updated CS conductor and coil design are presented. A CFX simulation carried out to check the pressure drop condition and a thermo-hydraulic analysis is performed. [ABSTRACT FROM AUTHOR]

• Superconducting toroidal field magnet of the Korean Fusion Demonstration Reactor is updated. • Smeared model of the K-DEMO TF magnet is applied to analyze magnetic field. • Structural analysis of the magnet structure is performed. • Thermo-hydraulic analysis for temperature margin during a current ramp up shows 1 K margin. After a conceptual design study for a steady-state Korean fusion demonstration reactor (K-DEMO) was initiated in 2012, the preliminary design of superconducting magnet system which is one of the key components of the K-DEMO has been done in 2015. The superconducting magnet system of K-DEMO has 16 Toroidal Field (TF) coils, 8 Central Solenoids (CS) modules and 12 Poloidal Field (PF) coils which use Cable-In-Conduit Conductors (CICC). For a high 7.4 T magnetic field generation at plasma center, high performance Nb3Sn based superconductor will be developed and used for TF magnet. Starting from the preliminary conceptual design, design updates of K-DEMO TF magnet is being conducted for checking engineering issues. For a validation of the magnet design, thermo-hydraulic stability and mechanical stability are being considered. In this work, the conceptual design studies for updates of K-DEMO TF magnet system is described. [ABSTRACT FROM AUTHOR]