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We open an avenue for discussing how we can pave the way for compliance with existing regulations is a far-reaching factor for settling nuclear fusion technology. Based on a model of the Korean Fusion Demonstration Reactor (K-DEMO) with a target fusion power of 2.2 GW, we assess the intrinsic safety determinants of internal energy sources, the expected radioactive waste, and the tritium management. Regarding these safety factors, we scrutinize the compatibility of the current legislative environment in Korea with K-DEMO and envisage foreseeable obstacles, such as licensing of the nuclear facilities and acceptability of the radioactive waste. Based on precedent licenses for the Korean Superconducting Tokamak Advanced Research (KSTAR) and lessons learned from the International Thermonuclear Experimental Reactor (ITER), we examine hazardous factors that would threaten regulatory compliance of K-DEMO. This approach can help shape a fusion-compatible framework for consolidating the necessary technical provisions and regulatory baselines reflecting social acceptance with a sense of safety. Fusion-compatible aspects in the regulatory environment are discussed, from fusion philosophy to subordinate administrative and technical guidelines, facility classification, and detailed methods guaranteeing integrity and safety. This paper will contribute to the timely settlement of fusion demonstration facilities and subsequent commercial plants.

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.

Heat load on the misaligned leading edges of tungsten castellated blocks based on tungsten (W), oxygen-free high conductive copper (OFHC-Cu), and copper-chrome-zirconium (CuCrZr) alloy are studied by COMSOL analysis and infrared (IR) measurements in KSTAR. IR measurements show that 1–3 MW/m 2 of heat flux has been deposited on the blocks during the inter-ELM (edge localized mode) phase in H-mode plasmas. COMSOL analysis indicates that the temperature of 1 mm leading edge in KSTAR under 3 MW/m 2 would reach the recrystallization temperature within 2 s and will be melted within 30 s during a long pulse H-mode shot. Rounded and double chamfered blocks show much better thermal response meaning that shaping of divertor block enhances the heat load handling capability. It seems that a simple COMSOL analysis describes heat load patterns on the tungsten blocks of different shapes qualitatively well. Therefore, simple analysis would be useful to make a quick prediction on heat load patterns of blocks with arbitrary shapes.

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.

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.

A pre-conceptual design study has been carried out for the Korean fusion demonstration reactor (K-DEMO) tokamak featured by high magnetic field (BT0 = 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 Li4SiO4 pebbles with Be12Ti 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.

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 Nb3Sn 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.

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.

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.

As a part of a conceptual design study for a steady-state demonstration reactor (K-DEMO) of which the major and minor radii are around 6.5 and 2.0 m, respectively, toroidal field magnets which can generate around 8 T at plasma center with a peak magnetic field of ~ 16 T are being studied. For conductor tests under such a high magnetic field, a new conductor test facility is required and a conceptual design activity is going on, in parallel. The conductor test facility, named SUCCEX (SUperConducting Conductor EXperiment), will feed the sample with a current up to 100 kA and the sample temperature will be varied from 4.5 K to ~ 20 K. Moreover, Sultan-like conductor tests, U-shaped joint-less sample tests will be conducted. The 16-T Nb3 Sn magnet, with a bore of ~ 1-m diameter, is divided into three concentric split pairs, inner coil (IC, peak field of 16.4 T), middle coil (MC, 13.6 T), and outer coil (OC, 7.9 T), connected in a series, with a nominal operating current of ~ 18 kA. Each split pair coil is made of its own distinctive rectangular shaped cable-in-conduit conductor (CICC). The overall design concept and preliminary results of electro-magnetic, structural, and thermo-hydraulic analysis are presented.

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 300 MWe and the competitiveness in cost of electricity (COE). The major radius is designed to be just below 6.5 m, 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.

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 6.5 m, considering practical engineering feasibilities. But still, high magnetic field at the plasma center around 8 T is expected to be achieved by using current state-of-the-art high performance Nb 3 Sn 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 300 MWe 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 300 MWe 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.

To analyse the experimental results of PF magnet during the commissioning test of KSTAR, a numerical model of the PF1 magnet and its CICC was developed for the evaluation of Joule heat from the AC loss due to the time varying field profile along the conductor during ramping-up or down taking into account the ferromagnetic effect of Incoloy 908 jacket. The calculation routines was carefully updated to make a proper estimation of the AC loss and applied to the estimation of AC loss parameters of the CICC of the KSTAR PF magnet.

According to the pre-qualification program for the production of Korean conductor for ITER TF magnet, a CICC(Cable In Conduit Conductor) sample for SULTAN test was developed based on the ?Option 2? specification which is the improved one among the two established baseline designs. For the assessment of performance, SULTAN test of the conductor sample was carried out in CRPP employing recently updated instrumentations on the sample and the joint of two conductor pieces, which has been discussed and modified for the proper estimation of . As a course of the analysis, the assessment of values was presented using the basic protocol of data reduction, and the discussion was made on the standard process of analysis. In addition, we investigated some exposed problems in the data manipulation such as non-linear voltage slopes being possible to be related to conductor characteristic itself and the initial offset problem in the calorimetric method. On the presentation of the test results, we also report the manufacture of the conductor with the qualification of the components including carefully designed Nb3Sn internal-tin superconducting strands for an improvement of conductor performance.

We have developed an apparatus to investigate the effect of transverse stress under axial strain on the critical current of superconducting strands. An internal tin Nb3Sn wire is soldered on a C-shape ring (called Pacman) made of 2% beryllium doped copper alloy and axial strain from -0.7 to 0.7% can be applied. The actual strain on a sample was estimated from a comparison with the critical current measurement results using a Walter spiral probe. Reversible transverse stress effects on the critical current were studied in this work and the transverse pressure was applied up to 40 MPa under ±0.5% axial strain. The critical current was initially increased about 3% as transverse load was applied and then decreased almost linearly under compressive axial strain. Similar behavior is observed under tensile axial strain but the critical current decreases rather sharply at higher load. A unified description based on 3 dimensional deviatoric strain was difficult for the sample studied in this work.

A revised specification of ITER TF conductor was proposed as "Option2" for the baseline design. The cabling pitches of early stages and the void fraction were adjusted in the consequence of previous R&D programs to improve the degradation of performance against cyclic current loadings in the magnetic field. To obtain the characteristics of the developed design, which are related to the effect of the electromagnetic load, the mechanical parameters and the contact resistances between components of cable were investigated employing a specially developed device of cryogenic press in the transverse direction for a short sample of the conductor cable. Through the measurements, the mechanical characteristics against the cyclic loads were discussed, as well as the effect on inter-strand resistances, according to the cable design.

A digital quench detection and protection system has been developed for a micro-sized superconducting magnetic energy storage system (SMES). The digital quench detection system based on a high-performance NI-6224 digital acquisition card accepts a standard set of input signal such as magnet induced voltage, current, pressure of cryostat. It also provides improved signal processing for sensitive and exact quench detection and data logging function. Once the superconducting magnet quenches, the quench detection system activates immediately the protection circuit. Some novel protection circuits are developed to dump the electromagnetic energy stored in the magnet; these greatly enhance the dump efficiency by maintaining the magnet terminal voltage constant during the dumping energy process. This paper detailedly describes the system and presents results.

A conceptual design study for Korean DEMO Reactor (K-DEMO) based on the Korean Fusion Energy Development Promotion Law (FEDPL) enacted in 2007 is now in progress. The K-DEMO at the first stage in our development plan is not regarded as a final DEMO but rather a test facility for a commercial reactor. The target of a preliminary blanket concept design study at the first stage is to promote the performance of breeding blanket in preference to any other. A preliminary blanket concept under consideration in the K-DEMO is based on a water-cooled ceramic blanket with lithium orthosilicate (Li 4 SiO 4 ) as breeder, beryllium (Be) as neutron multiplier, and reduced activation ferritic-martensitic (RAFM) steel as structural material. To investigate the possibility of utilization of heavy water (D 2 O) as a coolant of a breeding blanket for the K-DEMO, a comparative study of the two coolants, D 2 O and H 2 O has been conducted. Furthermore, the dependence of tritium breeding ratio (TBR) on volume ratios of water and breeder occupying a blanket, and the impact of D 2 O replacing H 2 O as a coolant on the K-DEMO blanket have also been investigated with the calculations on the basis of the Monte Carlo simulation MCNPX code. This paper presents the results of comparative study of the two coolants, D 2 O and H 2 O on our preliminary blanket concept for the K-DEMO.

For the goal of superconducting magnet applications in the advanced testing device for high temperature superconducting (HTS) wire and sample coils, a wide bore conduction-cooled superconducting magnet with available warm bore of phi 186 mm and center field of 5 T for the background magnetic field applications was designed and fabricated and tested. A sample cryostat with two GM cryocoolers is inserted in the background magnet. The system allows measurements to be performed in a repeatable and reliable fashion. The detailed design, fabrication and thermal analysis are presented in the paper.

The cryogenic system for the KSTAR superconducting (SC) magnets has been commissioned. It consists of a cold box, distribution boxes (DB) and cryogenic transfer lines. The cold box and DB #1 provide 600 g/s of supercritical helium to cool the SC magnets, their SC bus-lines, and the magnet support structures. It also provides 17.5 g/s of liquid helium to the current leads and supplies cold helium flow to the thermal shields. The main duties of the DB #2 are the relative distribution of the cryogenic helium among the cooling channels of each KSTAR cold component and the emergency release of over-pressurized helium during abnormal events such as quenches of the SC magnets. After individual commissioning, the system was integrated and cooled down with the KSTAR device. In this paper, the construction and commissioning results of the KSTAR cryogenic system will be introduced. In addition, we will present the cool-down results of the KSTAR device.

As an activity to build a new test facility for superconducting CICCs (Cable In Conduit Conductors), an insertion coil for an upgrade of the existing CS (central solenoid) model coil of KSTAR was designed and analysed to implement a superconducting magnet system being capable of pulsed operation up to 12 T. On the basis of a developed winding scheme of the insertion coil, some candidates of CICC design were proposed according to the result of an optimization method and a numerical code examining the thermo-hydraulic stability of CICC. After the design of winding pack and CICCs, we investigated thermo-hydraulic behaviors of the superconducting magnet with respect to given scenarios to estimate the limit of safe operation in pulsed mode.

Although there were many research activities for the development of superconducting Nb 3 Sn strands, the major one started under KSTAR (Korea Superconducting Tokamak Advanced Research) project in 1996. After the success of a large scale production test of Nb 3 Sn strand using the internal tin route, a new mass production facility is under operation since 2004. KAT (Kiswire Advanced Technology Ltd.), an affiliate of Kiswire Ltd., manufactured various types of Nb 3 Sn strands using the internal tin process optimized for fusion magnets. For the Nb 3 Sn strand of the KSTAR PF coil, each module has ∼190 niobium filaments and 19 modules are restacked for the strand production. For the ITER TF strand, there are two types of basic design. One of them has 37and 19 modules with 169–219 niobium filaments in each module. The other has 19 modules with 164–190 niobium filaments in each module. Both the designs satisfy the requirements for ITER TF strand with enough margins. The characterization of the strands is performed by hysteresis loss measurement, RRR (Residual Resistivity Ratio), n -value, and critical current density measurement vs. temperature, magnetic field, and strain. The critical current density of the strands reached around 1100 A/mm 2 at 12 T and 4.2 K. A well defined quality assurance program helped to produce a high quality strand with a piece length of more than 15 km. KAT has been provided Nb 3 Sn strand for KSTAR PF Coil and ready to produce the Nb 3 Sn strand for ITER TF coil. In this paper, the design concept, the fabrication procedure and the result of the strand performance test are discussed.

A conduction-cooled high temperature superconducting (HTS) magnet system through a solid nitrogen protection with energy storage of 30 kJ was developed. The HTS magnet system is used to investigate fast discharging performances with a constant output voltage. The superconducting magnet consists of 14 double pancakes wound with Bi2223 tape with the length of 200 m. The magnet has an outer diameter of 212 mm and a clear bore of 108 mm. Cryostat for the HTS magnet system is designed and the coil is cooled with a GM cryocooler together with the solid nitrogen protection technology. The superconducting magnet is fabricated and tested. The operating current is about 155 A with raping rate of 5 A/s. It can generate a central magnetic field of 4.31 T at a temperature lower than 20 K. In this paper, the magnet design, coil fabrication and cryogenic system are presented. Experimental research of the superconducting magnet as a constant voltage power supply was carried out.

A large bore NbTi superconducting magnet with liquid helium bath cooling has been developed for a micro-sized superconducting magnetic energy storage (SMES). The superconducting magnet has the clear bore of 480 mm, outer diameter of 550 mm and height of 500 mm. The magnet system is cooled in a compact cryostat with two GM cryocoolers. One of the cryocoolers is used to cool the thermal shielding. Another is used to re-condense the helium evaporation. A pair of high temperature superconducting current leads is used to reduce heat load to the cryostat. The operating current of the SMES system is rated at 325 A. Maximum energy storage is evaluated as 1.0 MJ at the rated operating current. Test results show that the stored energy can be dumped at the power of 7 kW to a dummy resistor load. This superconducting magnet system can be demonstrated as an Uninterruptible Power Supply (UPS). The superconducting magnet and the cryogenic system are described in this paper.

Temperature and strain responses of fiber Bragg grating sensors were measured in cryogenic environment. With temperature from room temperature down to 77 K or 4.2 K, the temperature response was found to be relatively linear above 100 K, and the temperature sensitivity decreases with the decrease of temperature, and to approximately zero for temperature less than 50 K. Technologies were developed to eliminate the multi-peaks in strain experiment, so the strain response was measured at 77 K, and was found to be linear at constant temperatures.

We have developed a probe for the critical current measurements of low temperature superconducting strands at various field, temperature and strain. A 30 cm-long strand sample is soldered on a Walters spiral made of beryllium copper alloy and compressive or tensile axial strain can be applied up to 0.7%. Temperature control ability was tested using a MgB2 wire up to 30 K. Three cernox sensors are attached on a spiral adjacent to the strand and the temperature of the spiral is controlled within plusmn50 mK of target temperature during the critical current measurement up to 80 A. Extensive critical current measurements have been carried out for an internal-tin processed Nb3Sn strand. It was found that the measured field, temperature and strain dependence of the critical current for the internal-tin Nb3Sn strand is in agreement with the recent scaling law based on strong coupling theory of superconductivity.

For the advanced testing of High Temperature Superconducting (HTS) wire/tape and sample coils, a wide bore conduction-cooled superconducting magnet with the available warm bore of 0186 mm and a center field of 5~6 T for background magnetic field applications was designed and fabricated. The magnet is composed of four coaxial coils. All the coils were connected in series and can be powered with a single power supply. The maximum magnetic field is 5.5 T. In order to support the high stress in superconducting magnet a detailed finite element (FE) analysis with electro-plastic model has been performed. A compact cryostat with a two-stage GM cryocooler was designed and manufactured for the magnet. The detailed design of magnet system is described in this paper.