Demonstration Fusion Reactor Research Articles

Comparative Study on the Reduction of Tritium Breeding Ratio Caused by Inventory Changes of a Solid-State Tritium Breeding Blanket in a Fusion Demonstration Reactor Using MCNP and FISPACT-II

The self-sufficient supply of tritium in the deuterium-tritium nuclear fusion demonstration reactor (DEMO) is a fundamental design requirement. But, it is hindered by depletion of tritium breeding materials resulting in reduction of tritium breeding ratio (TBR) less than the initial design value especially in the solid-state tritium breeding blanket (TBB) of the DEMO. Unlike the liquid tritium breeding blanket of DEMO, compensation measures of the depleted breeding material in the solid-state TBB will be its substitution depending on the reduction rate of TBR. To estimate the replacement period of the solid-state TBB, it is required to estimate the reduction rate of TBR according to the operation conditions of the DEMO and the physical configuration of a solid-state TBB. In this study, the representative simulation codes, MCNP and FISPACT-II, are used for assessment of the reduction rate of TBR with the benchmarking model which is modified from the one poloidal segment of the TBB in the Korean-DEMO. After 3 full power-year operations with the neutron irradiation with the benchmarking model, the TBR simulated by MCNP is reduced to 96.84% of the initially calculated TBR, but the TBR calculated by FISPACT-II is reduced to 90.57% of the initially calculated TBR.

Design update of DEMO BOP for HCPB BB concept with an energy storage system

An option for the Intermediate Coupling Design (ICD) of the European Demonstration Fusion Reactor (DEMO) Balance of Plant (BOP) configuration for Helium-Cooled Pebble Bed (HCPB) Breeding Blanket (BB) concept consists of the Primary Heat Transport System (PHTS), the Intermediate Heat Transfer System (IHTS), using HITEC salt as coolant, equipped with a thermal Energy Storage System (ESS), and the Power Conversion System (PCS). The aforementioned BOP configuration is capable of producing electricity continuously during the pulse time (2 h) and dwell time (10 min) of DEMO plant operation.The present study reports on the current (2023) thermal system design of DEMO HCPB ICD BOP for 12 operational states defined by the DEMO Design Central Team (DCT) in 2021. It was shown that the PCS is able to accommodate all of the 12 operational points defined based on the uncertainties of the energy map, provided that an additional bypass line at the outlet of the Divertor Cassettes (DIV-CAS) Heat Exchanger (HX) towards the low-pressure part of the feedwater line (to the condenser in this investigation) is installed in the system. This line is activated only in the cases, where the feedwater flow to the low temperature DEMO heat sources HXs is needed to be higher than that necessary to remove BB power at the Steam Generators (SGs). In all the other cases, the bypass line remains unused. Additionally, perspective improvements of DEMO HCPB BOP regarding thermal ESS, as well as PCS are discussed.

Exploratory neutronic evaluation on the enhancement of tritium breeding and energy multiplication capability using (Th,U)O2 in a solid-state tritium breeding blanket of a fusion demonstration reactor

The achievable tritium breeding ratio (TBR) in Korean fusion demonstration reactor (K-DEMO) that utilizing the deuterium-tritium nuclear fusion reaction for energy production is the critical design parameter for the self-sufficiency of tritium. However, the calculation uncertainty of the achievable TBR makes it difficult to fully understand the tritium fuel cycle in the K-DEMO. Moreover, if the achievable TBR is less than the required TBR, the calculation uncertainty of the achievable TBR makes the design limitation on the self-sufficient tritium supply of the K-DEMO. Therefore, making the achievable TBR larger than the required TBR is the most important goal of the K-DEMO design for the self-sufficient tritium supply. To create an achievable TBR with a margin greater than the required TBR, the modified neutron-multiplying material that is different from the typical neutron-multiplying material (9Be) is tested in this study. The blending of the fissionable (232Th) and fissile (233U) material with 9Be is attempted to enhance the neutron multiplication reaction resulting in a gradual increase of the achievable TBR according to the increase of the atomic faction of fissile material. With the addition of fissile material, it is found that energy multiplication accompanied by fission reaction is dramatic. The possibility of the self-breeding of 233U from 232Th is also justified. However, the neutronic damages resulting in the frequent replacement of the components in the tritium breeding blanket and generation of the high-level waste also occurred as the negative impact accompanied by the blending of 232Th and 233U with 9Be.

Conceptual design study of a large bore superconducting test facility magnet, SUCCEX

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.

The effect of Li2TiO3 pebbles on the mechanical properties of ARAA under breeding blanket conditions

The breeding blanket serves as a critical element for achieving a self-sustaining tritium cycle and generating sufficient energy in fusion reactors. This study focuses on candidate materials of the helium-cooled solid breeder blanket designed for a demonstration fusion reactor breeding blanket in Korea. We conducted compatibility tests between advanced reduced-activation alloy (ARAA) and Li2TiO3 pebbles at 550 °C for 28 days in He-0.1 vol% H2 purge gas conditions. Surface oxidation on ARAA was observed regardless of pebble presence. Double oxide films comprised of Fe-rich and Cr-rich layers formed on the surface of ARAA with pebbles. Lithium deposition was verified in both contact and non-contact areas of ARAA. Despite surface oxidation, the impact on tensile strength and elongation was found to be negligible.

Preliminary design and analysis activities of the deuteron accelerator target for tritium breeding unit test

A breeding blanket (BB) is a key component that multiplies neutrons and generates tritium in the fusion demonstration reactor. Evaluation of the performance and reliability of the BB is a very important engineering task, and it is going through the ITER TBM program. The Korean DEMO program needs to complement the ITER TBM program to validate the BB, and the test facility is necessary to validate the BB in the continuous wave operation. Based on this need, the Korea Institute of Fusion Energy has initiated a pre-conceptual study to verify the long-term performance and reliability of the BB.This study deals with neutronics and engineering analyses of a solid beryllium (Be) target that generates neutrons through a deuteron accelerator for the fusion neutron source. In the neutronics analysis, the yield and energy spectrum of neutrons generated according to the incident particles and energy were analyzed. As a result, in the case of 40 MeV deuteron, the yield of neutrons with 10∼20 MeV similar to the fusion neutron was the highest in the forward direction of 1.08E+15 n/s. In addition, vanadium (V) was selected as the blistering mitigation layer (BML), and the thickness of the Be target and the BML were optimized through neutronics analysis to minimize the blistering. In the engineering analysis, a preliminary conceptual design of the target to cool the nuclear heating caused by the nuclear reaction was performed. To improve the cooling capacity of the target composed of Be-V-CuCrZr with the pressurized water coolant, the design of the target was optimized and its thermal integrity was evaluated. Next, thermal stress was evaluated by coupling the temperature distribution of the target obtained through thermal analysis with structural analysis. Through this, the preliminary design concept of the target that satisfies thermal and mechanical integrity requirements was derived.

Design study of superconducting coil system for JA DEMO

A fusion demonstration (DEMO) reactor requires toroidal field (TF) coils larger than those used in ITER and can withstand higher electromagnetic forces. This creates significant challenges regarding the manufacturability of the TF coils, the increased electromagnetic forces on the TF inner leg case, and increased fabrication costs. This means that new winding pack designs and superconducting conductors must be designed to meet these requirements, and high-strength cryogenic steel for the coil case must be developed to withstand electromagnetic forces. This paper describes these possible solutions based on the Japanese DEMO (JA DEMO) design. In the layered winding of rectangular conductors, it was found that a new conductor geometry called the hybrid R-shape conductor can significantly reduce the stress on the turn insulation while reducing the superconducting wire to less than half. In addition, a guideline for designing the conductor strand structure was provided for a superconducting conductor subjected to electromagnetic forces 1.5 times greater than those of ITER. Furthermore, the effect of composition on strength was evaluated using several prototype materials based on high-Cr austenitic steels, and the feasibility of increasing the strength of cryogenic steels was confirmed.

Development of water-cooled cylindrical blanket in JA DEMO

The concept of the tritium breeding blanket for Japan’s DEMOnstration fusion reactor (JA DEMO) has been developed with pressure tightness against in-box loss-of-coolant accidents based on a water-cooled solid breeder concept. The cooling conditions are designed on the pressurized-water reactor water conditions which are the coolant temperature of 290 °C–325 °C and the operating pressure of 15.5 MPa, respectively. The point of the blanket design is to reduce the amount of structural material in casing as well as to ensure its pressure tightness. This is because a decrease in the amount of structural material improves tritium breeding ratio (TBR). A cylindrical structure, a thin wall casing structure which ensures pressure tightness and could increase TBR, is feasible. However, a relatively larger useless space is expected between modules when the cylindrical blanket modules are arranged in a vacuum vessel, which could decrease TBR. Therefore, the cylindrical blanket modules are to be in a close-packed arrangement to reduce useless space. The Be12Ti block (which shows a minor swelling compared to Be) is selected to achieve the target TBR as the net Be density is equivalent to the case of the Be pebble. The use of Be12Ti blocks can reduce or remove the cooling piping inside the module as the Be12Ti block has a higher thermal conductivity than pebbles. As a result of the neutronics, finite element method, and computational fluid dynamics analyses, it was found that the target TBR value can be achieved in cylindrical structure blanket that ensure pressure tightness.

Overview of Broader Approach activities

The Broader Approach (BA) activities aim to complement the ITER project and to promote the early realization of fusion energy through research, development, and tests of technologies supporting the future demonstration fusion reactor (DEMO). These activities are implemented under the BA agreement, which was signed and ratified in 2007 between Euratom and the Government of Japan. In essence, the BA activities consist of three projects:a) The Satellite Tokamak Programme Project JT-60SA, the world's largest superconducting tokamak until ITER starts, aims to support the assembly, commissioning and preliminary operation of ITER, and carry out demonstration and optimisation of steady-state operation of advanced plasma configurations for DEMO;b) The International Fusion Energy Research Centre (IFERC), which comprises three different sub-projects: the DEMO Design Research and Development Coordination Centre to coordinate design and R&D on materials and components for DEMO, the Fusion Computer Simulation Centre for the simulations of fusion plasmas, the analysis of experimental data, modelling of ITER operation, and contribution to the design of DEMO, and the Remote Experimentation Centre to allow scientists to participate remotely in fusion experiments from its control room in Japan;c) The Engineering Validation and Engineering Design Activities for the International Fusion Materials Irradiation Facility (IFMIF/EVEDA) focused on the design and validation of key components needed for the future Fusion Neutron Source facilities to characterize materials envisioned for DEMO.After 13 years of fruitful collaboration, Euratom and Japan launched the second phase of the BA activities in 2020. It is focused on exploiting and enhancing the facilities that have already been built and on working more closely than ever with ITER as the latter advances towards the first plasma. This paper will present a synthesis of the achievements already obtained thanks to the three BA projects and will address the future developments planned in 2023 and beyond.

Application of two-dimensional temperature response functions for reconstruction of divertor heat flux profile in commercial fusion reactors.

To keep the tritium breeding rate TBR > 1 and to meet the high heat load and neutron shielding requirements for the first wall and divertor in fusion demonstration (DEMO) reactors, the number of port plugs and other openings must be limited. To accomplish this, it is necessary to develop alternatives to the use of infrared (IR) thermography to determine the peak heat flux and the heat flux profile onto divertor targets. A divertor tile equipped with multiple temperature monitoring channels can be used to reproduce the temperature profile. To avoid the high temperatures and high neutron flux environment in a DEMO, the monitoring positions can be set well away from the irradiated surface. However, the spatial resolution of this method is lower than that provided by IR thermography. In the present work, we apply two-dimensional temperature response functions and the corresponding heat conduction model to temperature data obtained from a divertor tile surface in the large helical device to study the effects of the spatial resolution of the monitored temperature profile on the reconstructed heat flux profile. The findings provide information that will be useful in defining a method for embedding thermocouples into the divertor tiles of future DEMO reactors.

Estimation of impact of non‐local ion heat flux model on transport simulation for DEMO‐relevant scrape‐off layer plasma

AbstractIn a fusion DEMOnstration reactor (DEMO)‐relevant scrape‐off layer plasma (SOL), whose collisionality is lower than in the SOLs of present‐day tokamaks, kinetic effects are predicted to reduce the plasma heat conductivity along the magnetic field below the value obtained with the classical Spitzer–Harm model. As a part of ongoing efforts to improve the predictive capability of SOL heat transport calculations, we have implemented the non‐local heat flux model proposed by [Luciani, Mora and Virmont, Phys. Rev. Lett. 51 (1983) 1664–1667] (here referred to as the LMV model) in the integrated SOL–divertor simulation code SONIC in order to account for the kinetic effect on the heat conduction due to the parallel streaming of rapidly moving particles. Our transport simulations for the Japanese demonstration tokamak reactor concept, JA DEMO, show that the LMV model yields a significantly reduced ion parallel conductive heat flux density both on the low‐ and high‐field side of the upstream SOL. The heat flux obtained with the LMV model seems to be consistent with earlier kinetic simulations of tokamak ion transport. As a consequence of the reduced heat flux, a significant increase in the ion temperature and a decrease in the density have also been found over a broad area of the upstream SOL.

Scenario optimization for the tokamak ramp-down phase in RAPTOR: Part B. safe termination of DEMO plasmas

An optimized plasma current ramp-down strategy is critical for safe and fast termination of plasma discharges in a tokamak demonstration fusion reactor (DEMO), both in planned and emergency scenarios, avoiding plasma disruptions and excessive heat loads to the first wall. Plasma stability limits and machine-specific technical requirements constrain the stable envelope through which the plasma must be navigated. Large amounts of auxiliary heating are required throughout the ramp-down phase, to avoid a radiative collapse in the presence of intrinsic tungsten and seeded xenon impurities, as quantitatively estimated in this work. As the plasma current is reduced, the current density becomes increasingly peaked, reflected by a growing value of the internal inductance ℓi3 , resulting in reduced controllability of the vertical position of the plasma. The feasibility of different plasma current ramp-down rates is tested by applying an automated optimization framework embedding the RAPTOR core transport solver. Optimal time traces for plasma current Ip(t) and plasma elongation κ(t) are proposed, to satisfy an Ip -dependent upper limit on the plasma internal inductance, as obtained from vertical stability studies using the CREATE-NL code, as well as a constraint on the time evolution of q 95, to avoid an ideal MHD mode. A negative current density near the plasma edge is observed in our simulations, even for the most conservative Ip ramp-down rate, indicating significant transient dynamics due to a large resistive time.

Neutronics analysis for conceptual design of target system based on a deuteron accelerator-driven fusion neutron source

A Breeding Blanket (BB), producing tritium and heat energy converted to electricity, is a critical component for the successful Fusion Demonstration Reactor (DEMO). A dedicated neutron source is required to validate the performance of the DEMO BB in a continuous long-term fusion-like environment. Therefore, a pre-conceptual design study has been carried out for the Integrated Breeding Test Facility (IBTF) to evaluate the DEMO BB performance. The IBTF is based on a 40 MeV deuteron accelerator-driven system with a maximum current of 10 mA. A solid Beryllium (Be) target with a 20 cm width and height is used to irradiate the tritium breeding unit (TBU).This study conducted a neutronics analysis of the Be target system of the IBTF. The produced neutron energy spectrum and yield were evaluated on a beam energy of 30 to 50 MeV. The analysis results show that the forward direction neutron yield with the 40 MeV deuteron is the highest at 4.92E + 15 n/s, and the neutron ratio of 10–20 MeV is also the highest. A Blistering Mitigation Layer (BML) was added to prevent the blistering effect caused by the deuteron beam within the Be target. A comparison was done for the performance evaluation of the BML candidates vanadium, palladium, tantalum, and niobium. There were no significant differences in the forward direction neutron yield for the BML options. Vanadium was the most beneficial based on the activation analysis results. Therefore, vanadium with a 0.3 mm thickness was added behind the Be target with a 5 mm thickness. In addition, neutronics analysis was conducted for the test cell shielding designs. Concrete 3.5 m thick was required to satisfy a dose of less than 100 uSv/h during the beam operation. When evaluating the dose rate during maintenance, a dose along the cooling pipes was observed due to the activation of the pipes. The neutron flux with the cooling pipes was reduced by using B4C in the maintenance cell shielding, the dose rate was below 100 uSv/h in the test cell.Deuteron and neutron transport calculations were carried out using MCNP6 with the TENDL2019 cross-section library. Activity analysis and gamma source production were performed using FISPACT-Ⅱ with the TENDL2017 cross-section library.

T production using a high-temperature gas-cooled reactor for the DEMO fusion reactor: Li rod structure for the initial irradiation test

This work proposes tritium (T) production using a high-temperature gas-cooled reactor (HTGR) and studies it for the initial T inventory in a demonstration fusion reactor and a prior engineering test with T handling. At this stage, the aim is to investigate the compatibility between electricity and T production, which makes the stable confinement of T in the Li-loading rods during the HTGR operation period a crucial issue. The total T outflow into the helium gas during the reactor operation period was attempted to be reduced using Zr spheres with Ni coating as a T absorption material to avoid an increase in the inner T pressure. The basic hydrogen absorption properties of the Zr spheres with Ni coating, which coexist in an environment with Al2O3 and/or LiAlO2 oxides, were measured. The structures of the Li-loading rod for a typical commercial HTGR and irradiation test were studied using the obtained data. In addition, an outline of the irradiation test is reported.

Ab initio study of hydrogen in titanium beryllides

Titanium beryllide Be12Ti is a candidate material for the neutron multiplier for the demonstration fusion reactor DEMO. Experimental studies show that under certain conditions, Be12Ti may contain inclusions of other phases such as Be2Ti, Be17Ti2. In this regard, it is important to study the behavior of tritium and its isotopes in the crystal lattices of these phases. All calculations are performed using ab initio methods. Solution energies of a hydrogen atom in all non-equivalent interstitial sites of the three studied titanium beryllides were found to be lower than that in pure beryllium. The formation energy of all types of vacancies in all studied beryllides is found to be higher than that in beryllium. The binding energies of a single hydrogen atom located both inside and outside the vacancies are calculated. Hydrogen inside monovacancy is more strongly bound as compared to that outside this vacancy. It turned out that in some cases hydrogen can be captured by vacancy being outside of it. The results obtained are essential for further study of interstitial diffusion of hydrogen and analysis of tritium retention in titanium beryllides.

A New Device Concept of Magnetic Confinement Deuterium–Deuterium Fusion

A two-stage cascade magnetic compression scheme based on field reversed configuration plasma is proposed. The temperature and density of plasma before and after magnetic compression are analyzed. In addition, the suppression of the two-fluid effect and the finite Larmor radius effect on the tilting mode and the rotating mode of major magnetic hydrodynamic instability is studied, and finally, the key physical and engineering parameters of the deuterium–deuterium fusion pulse device are introduced. Further analysis shows that the fusion neutrons can be produced at an energy flux of more than 2 MW/m2 per year, which meets the material testing requirements for the fusion demonstration reactor (DEMO). If the recovery of magnetic field energy is taken into account, net energy outputs may be achieved, indicating that the scheme has a potential application prospect as a deuterium–deuterium pulse fusion energy.

Study of the Effective Torus Exhaust High Vacuum Pumping System Performance in the Inner Tritium Plant Loop of EU-DEMO

The requirement for a reduction of the tritium inventory of the European demonstration fusion reactor (EU-DEMO) has led to the active research and development of a continuously working pumping process termed “KALPUREX.” This process foresees the direct recycling of a large fraction of the unburnt hydrogen isotopologues via superpermeation in metal foil pumps during the burn phase. The remaining exhaust gas mixture is pumped by continuously operating, mercury-driven linear diffusion pumps. Diffusion pumps are kinetic high vacuum pumps whose pumping principle is based on the momentum transfer from a supersonic mercury vapor jet to the pumped gas mixture. Like many high vacuum pumps, they feature species-dependent pumping speeds. In the present work, we develop a simplified hybrid model of the high vacuum pumping train in order to estimate the effective pumping speed of the integrated system. The results of this model and its implications on the further development of the vacuum system are discussed for the burn and dwell phases of EU-DEMO.

Beamline and vacuum system design of HEBT for the A-FNS accelerator

The Advanced Fusion Neutron Source (A-FNS) project is progressing at the QST Rokkasho institute, Japan. The A-FNS will conduct fusion neutron irradiation tests of materials for a fusion demonstration reactor (DEMO reactor). For these tests a high intensity deuteron beam accelerator is used to produce a high neutron flux with the neutron energies around 14 MeV. The accelerator system accelerates a deuteron beam to a beam energy of 40 MeV, guides the beam towards a liquid lithium target and shapes the uniform beam at the Li target using a magnet system.Two of the main considerations for the A-FNS high energy beam transport (HEBT) design are protection of the Superconducting Radio-Frequency linear accelerator (SRF) and vacuum system design for the interface of the HEBT and the target system. For the A-FNS, back streaming of Li vapor from the windowless free surface of the Li target is a concern. In addition, widely different vacuum pressure must be maintained between the accelerator system and the target system. This paper shows the detailed design of the A-FNS HEBT section regarding these considerations.

Conceptual Design Study of a Test Facility Magnet for Fusion Conductor Samples

Design studies of the steady-state Korean fusion demonstration reactor (K-DEMO) magnet system have been conducted since 2013. The maximum magnetic field of the K-DEMO toroidal field (TF) coil is expected to be over 16 T. A high field superconducting magnet and a facility for testing the K-DEMO TF conductor samples are also being designed to prepare for the development of the K-DEMO TF conductor. To test straight fusion conductor samples compatible with the K-DEMO TF conductor, it is essential to have a large, 150 mm wide by 120 mm high sample space in the magnet. The required maximum magnetic field of the fusion conductor test magnet is expected to be over 15 T. The field consists of about a 15 T background field and some self-field. Both high current density Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn strands and high temperature superconductor (HTS) tapes are candidates for the superconducting materials for the magnet. In this work, two conceptual design study models of the high field magnet are introduced - a planar racetrack dipole based on HTS tape (I <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">c</sub> > 200 A/4-mm at 4.2 K, 18 T) cable-in-conduit conductor (CICC), and a non-planar flared end saddle dipole based on Nb <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sub> Sn strand (J <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">c</sub> > 2600 A/mm <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> at 4.2 K, 12 T) CICC. 2-dimensional magnetic and mechanical analyses of both magnet models were performed and the main features of the models are presented.

The investigation of the neutronic responses of components in the K-DEMO

The neutronic response of consisting materials of the Korean fusion DEMOnstration reactor (K-DEMO) in its fusion neutron environment is a crucial consideration factor from the conceptual design stage of the K-DEMO. Especially in the design of in-vessel components (IVCs) of the K-DEMO that are placed in the most extreme neutron irradiation field, neutronic damage of constituent components is a major limiting factor that determines the lifetime of IVCs. From the analysis of the neutronic response of IVCs of the K-DEMO, the neutron wall loading (NWL) related to the tritium breeding ratio and nuclear heating of IVCs can be quantified to assess the self-sufficient supply of tritium, and thermal energy transferred from fusion neutrons, respectively. The calculated NWL shows that the harmonizing design of the cooling configuration of each blanket segment with the corresponding NWL is critical thermos-hydraulic design issue for the efficient utilization of thermal energy in the blanket. Another finding is that the double null magnetic field configuration and related blanket configuration with a water-cooled pebble bed of the K-DEMO make a self-sufficient tritium supply challenging. The implicated lifetime of the first plasma-facing tungsten wall of the K-DEMO is around 2 full power years (FPY) in the severely neutron-irradiated region. For the reduced activation ferritic/martensitic steel layers in the blanket, the lifetime of it is estimated around 4 FPY in the inboard region. Based on the response analysis results of this study, optimization of the design of the K-DEMO will continue iteratively in the future.