System analysis study for Korean fusion DEMO reactor

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 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 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.     

Conceptual design study of fusion DEMO plant at SWIP

The basic definition and development strategy of the DEMO plant based on the Chinese fusion power plant (FPP) program are presented briefly. A conceptual design study of fusion HCSB-DEMO reactor with a fusion power of 2550 MW and a neutron wall loading of 2.3 MW/m² is performed recently. Three sets parameters of core plasma for different DEMO design objectives are proposed. A helium-cooled blanket system with ceramic breeder (Li₄SiO₄), the structure material of low-activation ferritic steel (LAF/M) and Be neutron multiplier based on Chinese ITER HCSB-TBM design foundation are considered. The design parameters, preliminary analyses and the basic structure as well as development strategy of HCSB-DEMO reactor are introduced.     

A preliminary conceptual design study for Korean fusion DEMO reactor

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.     

Contribution to DEMO reactor RH maintenance assessment

The scope of this paper is a preliminary assessment of the maintenance scheme in support of the European study for the next generation of fusion reactor: DEMO. Despite other fusion machine requiring remote handling maintenance operations, DEMO is supposed to work under steady state operational conditions. Therefore, requirement on the maintenance scheme is stronger. To target a good availability of the machine along machine operation plan, it is necessary to draw an adequate maintenance scheme. Indeed, due to the high fluxes generated by the plasma in the vacuum vessel, the first wall lifetime is limited, so the frequent replacement is necessary. On current fusion experimental machine, as first wall load conditions are less severe, DEMO condition implies high level of requirement on maintenance time. During DEMO lifetime, several full first wall replacements are expected. To provide access to the vacuum vessel machine for first wall removal, preparatory work is required to set the machine to adequate maintenance conditions and to open the machine properly, the same situation at the end of the maintenance period. Shutdown duration for first wall replacement should be as short as possible to reach the availability target of the machine. From this statement, the maintenance duration should not exceed 20% of the total lifetime of the reactor operation. First wall segmentation (i.e. total number of component to replace) has a high impact onto the replacement time. Considering the number of feasible designs for the first wall segmentation, we concentrate remote handling concept assessments one type of segmentation, the one minimizing the numbers of modules to replace [4], [5], [6]. Assumption on Divertor segmentation for these DEMO studies have similarities with Divertor ITER design; therefore ITER design output is relevant [1], [2]. We assume divertor removal performed in shadow time, while removing the other first wall modules.     

Conceptual design for the superconducting magnet system of a pulsed DEMO reactor

A methodology has been developed to consistently investigate, taking into account main reactor components, possible magnet solutions for a pulsed fusion reactor aiming at a large solenoid flux swing duration within the 2–3 h range. In a conceptual approach, investigations are carried out in the equatorial plane, taking into account the radial extension of the blanket-shielding zone, of the toroidal field magnet system inner leg and of the central solenoid for estimation of the pulsed swing.Design criteria are presented for the radial extension of the superconducting magnets, which is mostly driven by the structures (casings and conductor jacket). Typical available cable current densities are presented as a function of the magnetic field and of the temperature margin.The magnet design criteria have been integrated into SYCOMORE, a code for reactor modeling presently in development at CEA/IRFM in Cadarache, using the tools of the EFDA Integrated Tokamak Modeling task force.Possible solutions are investigated for a 2 GW fusion power reactor with different aspect ratios.The final adjustment of the DEMO pulsed reactor parameters will have to be consistently done, considering all reactor components, when the final goals of the machine will be completely clarified.     

聚变发电反应堆概念设计研究

在广泛分析聚变能相关领域研究发展状况和国际热核聚变实验堆(ITER)物理与技术基础上,提出了一个考虑了技术可行性的聚变发电反应堆概念(称之为FDS Ⅱ)。这个概念具有ITER参数适量外推的等离子体物理与技术水平的聚变堆芯和具有发展潜力的液态锂铅氚增殖包层,在对这个概念进行中子学、热工水力学、力学、安全与环境影响和经济学等一系列计算分析的基础上,给出了初步的概念设计和进一步设计优化的共性原则。     

Study of safety features and accident scenarios in a fusion DEMO reactor

After the Fukushima Dai-ichi nuclear accident, a need for assuring safety of fusion energy has grown in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of Broader Approach DEMO Design Activities (BA-DDA). This paper reports progress in the fusion DEMO safety research conducted under BA-DDA. Safety requirements and evaluation guidelines have been, first of all, established based on those established in the Japanese ITER site invitation activities. The radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO concept. This concept employs in-vessel components that are cooled by pressurized water and built of a low activation ferritic steel (F82H), contains solid pebble beds made of lithium-titanate (Li₂TiO₃) and beryllium–titanium (Be₁₂Ti) for tritium breeding and neutron multiplication, respectively. It is shown that unlike the energies expected in ITER, the enthalpy in the first wall/blanket cooling loops is large compared to the other energies expected in the reference DEMO concept. Reference accident event sequences in the reference DEMO in this study have been analyzed based on the Master Logic Diagram and Functional Failure Mode and Effect Analysis techniques. Accident events of particular concern in the DEMO have been selected based on the event sequence analysis and the hazard assessment.     

Neutronic design analyses for a dual-coolant blanket concept: Optimization for a fusion reactor DEMO

The generation of design specifications for a DEMO reactor, including breeding blanket (BB), vacuum vessel (VV) and magnetic field coils (MFC), requires a consistent neutronic optimization of structures between plasma and MFC. This work targets iteratively to generate these neutronic specifications for a Dual-Coolant He/Pb15.7Li breeding blanket design. The iteration process focuses on the optimization of allowable space between plasma scrapped-off-layer and VV in order to generate a MFC/VV/BB/plasma sustainable configuration with minimum global system volumes. Two VV designs have been considered: (1) a double-walled option with light-weight stiffeners and (2) a thick massive one. The optimization process also involves VV materials, looking to warrant radiation impact operational limits on the MFC. The resulting nuclear responses: peak nuclear heating in toroidal field (TF) coil, tritium breeding ratio (TBR), power amplification factor and helium production in the structural material are provided.     

Concept for a vertical maintenance remote handling system for multi module blanket segments in DEMO

The anticipated high neutron flux, and the consequent damage to plasma-facing components in DEMO, results in the need to regularly replace the tritium breeding and radiation shielding blanket. The current European multi module segment (MMS) blanket concept favours a less invasive small port entry maintenance system over large sector transport concepts, because of the reduced impact on other tokamak systems – particularly the magnetic coils. This paper presents a novel conceptual remote maintenance strategy for a Vertical Maintenance Scheme DEMO, incorporating substantiated designs for an in-vessel mover, to detach and attach the blanket segments, and cask-housed vertical maintenance devices to open and close access ports, cut and join service connections, and extract blanket segments from the vessel. In addition, a conceptual architectural model for DEMO was generated to capture functional and spatial interfaces between the remote maintenance equipment and other systems. Areas of further study are identified in order to comprehensively establish the feasibility of the proposed maintenance system.     

Evaluation of remote maintenance schemes by plasma equilibrium analysis in Tokamak DEMO reactor

The remote maintenance schemes in a DEMO reactor are categorized by insertion direction, blanket segmentation, and divertor maintenance scheme, and are quantitatively evaluated by analysing the plasma equilibrium. The positions of the poloidal field (PF) coil are limited by the size of the toroidal field (TF) coil and the maintenance port layout of each remote maintenance scheme. Because the PF coils are located near the larger TF coil and far from the plasma surface, the horizontal sector transport maintenance scheme requires the largest part of total PF coil current, 25% larger than that required for separated sector transport using vertical maintenance ports with segmented divertor maintenance (SDM). In the unsegmented divertor maintenance (UDM) scheme, the total magnetic stored energy in the PF coils at plasma equilibrium is about 30% larger than that stored in the SDM scheme, but the time required for removal and installation of all the divertor cassettes in the UDM scheme is roughly a third of that required in the SDM scheme because the number of divertor cassettes in the UDM scheme is a third of that in the SDM scheme. From the viewpoint of simple maintenance operations, the merit of the UDM scheme has more merit than the SDM scheme.     

Research and development status on fusion DEMO reactor design under the Broader Approach

The main objective of DEMO design activity under the Broader Approach is to develop pre-conceptual design of DEMO options by addressing key design issues on physics, technology and system engineering. This paper describes the latest results of the design activity, including DEMO parameter study, divertor and remote maintenance. DEMO parameter study has recently started with “pulsed” DEMO having a major radius (R_p) of 9 m, and “steady state” DEMO of R_p = 8.2 m or more. Divertor design study has focused on a computer simulation of fully detached plasma under DEMO divertor conditions and the assessment of advanced divertor configuration such as super-X. Comparative study of various maintenance schemes for DEMO and narrowing down the schemes is in progress.     

Parametric neutronic analysis of HCLL blanket for DEMO fusion reactor utilizing vacuum vessel ITER FDR design

Three-dimensional parametric neutronics calculations using the Monte Carlo code MCNP-4C have been performed for a DEMO-type reactor based on the Helium-Cooled Lithium-Lead (HCLL) blanket. The aim of the analysis was to minimize the radial blanket thickness, while ensuring tritium self-sufficiency and to assess the shielding performance of the reactor in terms of the radiation loads to the super-conducting toroidal field (TF) coils. It was found that tritium self-sufficiency can be achieved with a breeder zone thickness reduced to no more than 55 cm at a ⁶Li enrichment of 90%. Assuming a ⁶Li enrichment of 60%, a breeder zone thickness of 60 cm is required to achieve the target TBR of 1.10 which is assumed to be sufficient to cover potential tritium losses and uncertainties. With regard to the shielding performance it was found that the design limits for the radiation loads to the TF-coil can be met with radial blanket thicknesses of 75 cm, 60 cm and 55 cm utilizing a two-component shield of Eurofer steel and tungsten carbide between the breeder zone and the vacuum vessel. The blanket variants with larger radial breeder zone show better shielding performances due to the reduced Eurofer shielding material acting as gamma radiation emitter in between the breeder zone and the vacuum vessel. In particular the radiation dose absorbed in the Epoxy insulator was shown to be the most critical quantity in this regard.     

A low wall-loading DEMO reactor design with high priority for early and reliable realization of a tokamak fusion reactor over the cost performance

Based on scientific databases adopted for designing ITER plasmas and on the advancement of fusion nuclear technology from the recent R&D program, a low wall-loading DEMO fusion reactor has been designed, where high priority has been given to the early and reliable realization of a tokamak fusion plasma over the cost performance. Since the major radius of this DEMO reactor is chosen to be 10 m, plasma ignition is achievable with a low fusion power of 0.8 GW and an operation period of 4–5 hours is available only with inductive current drive. The low ignition power makes it possible to adopt a first wall with an austenitic stainless steel, for which significant databases and operating experience exists, due to its use in the presence of neutron irradiation in fission reactors. In step with development of advanced materials, a step-wise increase of the fusion power seems to be feasible and realistic, because this DEMO reactor has the potential to produce a fusion power of 5 GW.     

Pre-conceptual design assessment of DEMO remote maintenance

EDFA, as part of the Power Plant Physics and Technology programme, has been working on the pre-conceptual design of a Demonstration Power Plant (DEMO). As part of this programme, a review of the remote maintenance strategy considered maintenance solutions compatible with expected environmental conditions, whilst showing potential for meeting the plant availability targets. A key finding was that, for practical purposes, the expected radiation levels prohibit the use of complex remote handling operations to replace the first wall. In 2012/2013, these remote maintenance activities were further extended, providing an insight into the requirements, constraints and challenges. In particular, the assessment of blanket and divertor maintenance, in light of the expected radiation conditions and availability, has elaborated the need for a very different approach from that of ITER. This activity has produced some very informative virtual reality simulations of the blanket segments and pipe removal that are exceptionally valuable in communicating the complexity and scale of the required operations. Through these simulations, estimates of the maintenance task durations have been possible demonstrating that a full replacement of the blankets within 6 months could be achieved. The design of the first wall, including the need to use sacrificial limiters must still be investigated. In support of the maintenance operations, a first indication of the requirements of an Active Maintenance Facility (AMF) has been elaborated.     

Study on containment safety systems for Korean fusion DEMO plant

The purpose of this study is to review the strategy for radiation barriers in the fusion power plants and to produce simulation data for the conceptual design of safety features to maintain the integrity of such barriers as a part of R&D program through the National Fusion Research Institute of Korea. Even though the amount of radioactive source term in fusion power plants should be much less than that of fission power plants, internal as well as external events can result in damage to facilities such that public can be critically exposed by radiation. In the first part of this study, we reviewed and compared the multiple defenses to protect radioactive hazard in fission and fusion power plants. Containment was characterized as an indispensable physical barrier and the integrity of containment particularly enveloping a fuel cycle which is a major radioactive source term, tritium, should be secured. Since water is assumed as one of the coolant options in the Korean fusion DEMO plant, the thermo-hydraulic analysis was carried out using computer simulations to produce key parameters related with the integrity of containment in the second part. The performance of both of active and passive safety features to control the key parameters was compared to take recent fission technologies into account.     

聚变高温制氢反应堆概念设计研究

在深入分析相关领域研究发展状况的基础上,提出了具有较好技术可行性的聚变高温制氢反应堆概念(称之为FDS-Ⅲ),包括具有先进等离子体物理和技术水平的聚变堆芯、先进高温锂铅包层(HTL)、可减少热流分布密度的"垂直靶板"偏滤器以及相应的功率转换系统。尤其是提出了HTL包层新概念,其特点是选用技术基础相对成熟的低活化铁素体/马氏体钢作结构材料,在锂铅流道中使用可耐高温的多层流道插件,实现约1000℃的出口温度,可应用于制氢。初步性能分析表明FDS-Ⅲ制氢堆及其包层概念具有较好的技术可行性。     

Design failure mode and effect analysis for Korean fusion DEMO plant

The construction of Korean Fusion DEMO Plant (KFDP) for a demonstration of the plasma analysis and engineering feasibilities is planned in 2030s based on the Korean fusion technology roadmap. The radiation safety should be assured for nuclear facilities, so that, the KFDP is required to research for the regulatory requirements and industrial codes and standards. The final design guidance of the engineered safety features should be served in future. As the first step for this research, the failure modes and effects analysis in a design stage was performed. This leads to find the list of potential hazard elements and to obtain the list of initiating events for the future probabilistic risk assessment. The hazard elements expected to seriously threaten the integrity of the KFDP were investigated and determined to quantify the effect of the initiating events in the effect analysis: (1) a total loss of active cooling water to occur during the burn with decay heat calculation and (2) coolant ingress from cooling circuits into the vacuum vessel, cryostat and containment building. For those initiating events, the quantitative simulations using transient mass and energy calculation and computational fluid dynamics were performed.     

Effects of lithium burn-up on TBR in DEMO reactor SlimCS

Lithium in a breeding blanket is burned up through neutron nuclear reactions in fusion DEMO reactors. Effects of decrease of solid breeder materials due to lithium burn-up on tritium breeding ratio (TBR) are not systematically calculated in the past. For the SlimCS blanket design, TBR is calculated taking into account the lithium burn-ups by one dimensional Sn radiation transport calculation code ANISN in this study. The ⁶Li burn-ups are 8–79% after 10-year operation. TBR due to ⁶Li decreases to 40% of the initial one in some layer, while it increases in some layers. The TBR integrated over all the blanket decreases to around 96% of the initial one. The study makes it clear that the reduction of the TBR due to the lithium burn-up is not so large.     

An exploratory study on the gaps and pathways to the Korean fusion DEMO

With the vision of being an early demonstrator of fusion energy, the strategic plans for the Fusion DEMO program of Korea (K-DEMO program) has been developed. A staged development of the K-DEMO plant was considered in the strategic plans as to verify technical feasibility in the first stage and economic feasibility in the second stage. The top-tier design requirements and assumptions of the first stage K-DEMO plant are defined and postulated. With these requirements and assumptions, the desired and current status of nuclear fusion technologies are compared to identify the gaps to be filled to design, fabricate, construct, and operate it. The pathways from KSTAR, ITER to K-DEMO plant have also been studied to identify R&D activities for K-DEMO program that are to go in parallel with KSTAR and ITER are extracted from the pathways. Cross-cutting with the fusion R&D activities of the other countries and utilizing the commonalities with the existing systems are discussed with the provision of open-innovation strategy that is one of the key strategies of K-DEMO program. The priority of the R&D activities of K-DEMO program is qualitatively determined in consideration of the gaps, cross-cutting, and risks associated with the R&D investments.