DEMO Reactor Research Articles

WASTE CLASSIFICATION ASSESSMENT OF NUCLEAR STEELS FOR FUSION POWER APPLICATIONS

Fusion power is an attractive option for the world’s future energy needs. An important goal for fusion is to avoid the severe radioactive waste issues associated with nuclear fission. However, the neutrons produced in the fusion plasma reaction impinge on the surrounding reactor structure causing nuclear activation. It is hoped that activated material from fusion facilities can be disposed of as low level waste 50-100 years after operation ceases, but recent work suggests this may be difficult to achieve. This work presents inventory simulations for a number of potential fusion steels, for two neutron irradiation conditions typical of the DEMO reactor concept. The results are used to determine if the steels meet low level waste regulations, for a number of different international waste management systems. These results show that steels do not appear able to consistently meet low level waste requirements when exposed to near-plasma neutron fluxes. They have more success when exposed to lower fluxes, but traditional steels will still struggle to meet low level waste requirements in a fusion environment.

ОШИБКИ В РАДИАЦИОННОМ НАГРЕВЕ УРАНА ПРИ РАСЧЕТАХ С ИСПОЛЬЗОВАНИЕМ БИБЛИОТЕКИ FENDL 3.1D

The work is devoted to the analysis of uranium radiation heating results under the action of neutron radiation. The revealed results discrepancies in the radiation heating modeling of the uranium 235 and 238 using FENDL 3.1d and ENDF nuclear data libraries motivate the study. At the same time, all initial data in the computational model, namely, the neutrons energy spectrum, mass of fissile material, geometry and surrounding structures composition were identical. In the course of the work, the results of energy release calculations in uranium cells were compared using various libraries. Analytical evaluation of the radiation heating results in test cells is given. The obtained data indicate erroneous values of neutron kerma factors in FENDL 3.1d nuclear data library. Fusion Evaluated Nuclear Data Library was created by the working group of the Nuclear Data Section of the IAEA in the mid-1990s. FENDL is focused on application in the design of fusion plants as well as plants for materials testing such as IFMIF. The expanded application required the addition of some materials to the library, the expansion of the energy range to higher energies and the addition of data for reactions caused by charged particles. Version 3.1d was released in January 2018 and is widely used in nuclear analysis of structural materials for the experimental fusion reactor ITER and the demonstration fusion reactor DEMO. The results of this work have been submitted to the ITER International Organization contact group for presentation to the developers of the FENDL library.

Simulation studies of divertor detachment and critical power exhaust parameters for Japanese DEMO design

• Simulation study for the divertor design of Japanese fusion DEMO reactor. • Systematic scans of critical power exhaust parameters for the divertor operation. • Heat load and detachment plasma profiles were evaluated in the low density condition. • Restrictions of the divertor operation under more severe conditions were determined. Handling of a large thermal power exhausted from the confined plasma is one of the most important issues for ITER and DEMO. A conventional divertor, which has the closed geometry similar to that of ITER and longer leg of 1.6 m, was proposed for the Japanese (JA) DEMO reactor ( R p / a p = 8.5/2.42 m). A radiative cooling scenario of Ar impurity seeding and the divertor performance have been demonstrated by SONIC simulation, in order to evaluate the power exhaust in JA-DEMO 2014 (primary design with P sep ~ 283 MW) and JA-DEMO with higher plasma elongation (a revised design with P sep ~ 235 MW). The divertor operation with the peak q target ≤ 10 MWm −2 was determined in the low n e sep of 2–3 × 10 19 m −3 under the severe conditions of reducing radiation loss fraction, i.e. f* rad div = ( P rad sol + P rad div )/ P sep , and diffusion coefficients ( χ and D) . The divertor geometry and reference key parameters ( f* rad div ~ 0.8, χ = 1 m 2 /s and D = 0.3 m 2 /s) were so far consistent with the power exhaust concepts in the n e sep range, and the revised JA-DEMO design has advantages of wider n e sep range and enough margin for the divertor operation. For either severe assumption of f* rad div ~ 0.7 or χ and D to the half value, higher n e sep operation was required for the primary design in order to control the peak q target ≤ 10 MWm −2 , i.e. the operation window was reduced. Applying the two severe assumptions, the divertor operation was difficult in the low n e sep range for the both designs.

Preparations for a European R&D roadmap for an inertial fusion demo reactor.

A European consortium of 15 laboratories across nine nations have worked together under the EUROFusion Enabling Research grants for the past decade with three principle objectives. These are: (a) investigating obstacles to ignition on megaJoule-class laser facilities; (b) investigating novel alternative approaches to ignition, including basic studies for fast ignition (both electron and ion-driven), auxiliary heating, shock ignition, etc.; and (c) developing technologies that will be required in the future for a fusion reactor. A brief overview of these activities, presented here, along with new calculations relates the concept of auxiliary heating of inertial fusion targets, and provides possible future directions of research and development for the updated European Roadmap that is due at the end of 2020.This article is part of a discussion meeting issue ‘Prospects for high gain inertial fusion energy (part 2)’.

Influence of Krypton Seeding on EU DEMO Operation with Lithium Divertor

DEMO reactor with liquid lithium divertor operation in presence of krypton seeding is analyzed. Integrated core-scrape-off layer-divertor code COREDIV is used. Sputtering model is revised and sputtering dependence on the divertor surface temperature fitted to the experimental data is added. It was found that Kr seeding is beneficial for DEMO operation. Plasma fuel dilution by lithium ions is reduced to the level of 1% for high level of Kr seeding and fusion power is comparable to the modelling results for argon or neon seeded DEMO with tungsten divertor. Simulations show that high confinement mode is possible in presence of Kr seeding (power across the separatrix stays above the L-H threshold).

High-heat-flux technologies for the European demo divertor targets: State-of-the-art and a review of the latest testing campaign

Divertor target is one of the most critical in-vessel components in a fusion power plant being in charge of particle and power exhaust. The targets are exposed to severe thermal loads produced by steady bombardment of impinging plasma flux. Since 2014, integrated R&D efforts have been continued aiming at developing a design concept and high-heat-flux (HHF) technologies for divertor targets of the European DEMO reactor. Recently, the second round (2017–2019) of the R&D program has been concluded. As in the first R&D round, five water-cooled target design concepts were further developed and evaluated. Fabrication technologies were improved reaching a consolidated production quality. Extensive HHF tests were conducted using small-scale mock-ups for extended loading regimes (heat flux: 20–32MW/m²). Comparative studies were performed to investigate effects of copper interlayer thickness (0.1–1 mm) and different tungsten armor materials. In the present paper, the final results of the second round HHF testing campaign are reported. The HHF performance of each design variant is discussed based on in-situ diagnostic data (infrared thermography), ultrasonic inspection images and postmortem metallographic micrographs. All monoblock-type design concepts passed the specified qualification criterion (≥500 pulses at 20 MW/m², coolant: 130 °C) without any failure or armor cracking. Moreover, two of them (ITER-like and composite pipe) remained fully intact even under 25 MW/m² (100 pulses) and 32 MW/m² (5 pulses).

System Studies on the Fusion-Fission Hybrid Systems and Its Fuel Cycle

This paper is devoted to applications of fusion-fission hybrid systems (FFHS) as a powerful neutron source implementing transmutation of minor actinides (MA: Np, Am, Cm) extracted from the spent nuclear fuel (SNF) of nuclear reactors. Calculations which simulated nuclide kinetics for the metallic fuel containing MA and neutron transport were performed for particular facilities. Three FFHS with fusion power equal to 40 MW are considered in this study: demo, pilot-industrial and industrial reactors. In addition, needs for a fleet of such reactors are assessed as well as future FFHSs’ impact on Russian Nuclear Power System. A system analysis of nuclear energy development in Russia was also performed with the participation of the FFHSs, with the help of the model created at AO “Proryv”. The quantity of MA that would be produced and transmuted in this scenario is estimated. This research shows that by the means of only one hybrid facility it is possible to reduce by 2130 the mass of MA in the Russian power system by about 28%. In the case of the absence of partitioning and transmutation of MA from SNF, 287 t of MA will accumulate in the Russian power system by 2130.

Verification and Validation of the Geant4 Monte Carlo Code Toolkit for DEMO TBR Evaluations

Verification and Validation analyses of the Monte Carlo code toolkit Geant4 were conducted with nuclear data from the JEFF-3.3 library to show the suitability for fusion applications. We present benchmarking against data of the engineering design oriented HCPB (Helium Cooled Pebble Bed) Tritium Breeder Module (TBM) mock-up experiment taken from the SINBAD compilation and nuclear analyses in comparison with MCNP of the DEMO reactor equipped with a HCPB breeding blanket. Missing functionalities for Geant4 were developed such as a track length detector with multiple energy bins and tally multiplication, reflective boundaries, and statistical checks in a post-processing step. The activation foil measurements in the TBM mock-up revealed Geant4 underestimates the uncollided and fast neutron flux increasingly with penetration depth, while being consistent with MCNP for the low energy neutron flux. Geant4 underestimates the experimental tritium activity in the breeder pellets by up to 2.7% more than MCNP, which is within the experimental uncertainty. The DEMO analyses yielded a 0.97% lower Tritium Breeding Ratio (TBR) of Geant4 than MCNP for a homogenized breeder blanket, and a 1.46% higher TBR of Geant4 than MCNP for a mostly heterogenic blanket. This good agreement with MCNP demonstrates Geant4’s suitability for DEMO TBR evaluations.

Development and characterization of advanced neutron multiplier materials

Neutron multiplier materials are essential for self-sufficient tritium production and are closing the fuel cycle of fusion reactors. Until now, a concept of pebble bed consisting of interchanging layers of beryllium and lithium ceramic pebbles was considered for the Helium-Cooled Pebble Bed (HCPB) tritium-breeding module of the first experimental fusion reactor ITER as well as for the next demonstration fusion reactor DEMO. However, this concept depends on the availability of large amounts of pure beryllium pebbles and is also limited by its material properties like for example the tritium accumulation under irradiation.The results of tritium retention and analytical microstructural studies of beryllium pebbles obtained within the framework of the HIDOBE irradiation campaign suggest that a significant fraction of generated tritium (up to 100% below 500°C) is trapped within helium bubbles. Being negligible in the ITER tritium-breeding module (TBM), the total accumulated tritium inventory imposes severe safety issues and exceeds acceptable limits for the DEMO blanket. Therefore, advanced neutron multiplier materials such as beryllides have to be well characterized for their applicability in the HCPB blanket of DEMO and beyond. The usage of an advanced material with lower volumetric swelling, lower tritium retention, increased irradiation and chemical resistance as well as with higher melting temperature allows to switch from the pebble bed concept to a solid hexagonal block-based one. For the fabrication of titanium beryllide samples both the semi-industrial fabrication route utilizing the hot extrusion of rods and the industrial approach using vacuum hot pressing of a Be-Ti powder mixture were explored. In this contribution, we discuss reasons for the transition from pure beryllium to beryllides, respective changes of the HCPB blanket design and a successful demonstration of the feasibility of a beryllide block fabrication by an industrial method.

Irradiation damages of structural materials under different irradiation environments

For the advancement and development of nuclear systems used in heavy irradiation environments such as fusion DEMO reactors, fission reactors, fast reactors, and accelerator driven target systems, it is necessary to fully understand the changes of mechanical properties and the other properties of the materials induced by irradiation and to clarify the synergistic effect of displacement damage and helium generation. In this study the mechanical property changes and microstructural development induce by displacement damage and helium production have been mainly examined in austenitic stainless steels, 316FR and type304, and ferritic/martensitic steel, HCM12A, irradiated at around 550oC in JRR-3M reactor and/or JOYO fast reactor. At 550°C, 316FR steel was superior to 304 steel in terms of the amount of ductility and strength with respect to irradiation resistance. It is noteworthy that at 550 °C as well as room temperature, the higher fracture strength of the 316FR steels is a remarkable result. It is found that helium atoms strongly influenced on creep lifetime of the irradiated austenitic stainless steel, 316FR. It was found that the ratio of creep rupture time is slightly lower than the lower limit of previous study's Miyaji and co-workers in the region from 0.01 appm to 1 appm. It is also found that the lower limit of reduction ratio of creep rupture time (irradiation specimen to unirradiation one) does not decrease linearly with the helium production above 10 appm up to about 33 appm. dpa enhanced the reduction of creep lifetime. Recent R&D of high-energy accelerator driven target systems used under heavy irradiation environment is also introduced and discussed for high radiation resistance materials such as Ti alloys with very high number density of nano size precipitate, which have been studying under RaDIATE collaboration, from points of view of irradiation damage and materials development.

Influence of Lithium Mass Transfer on Tritium Behavior in Pebbles of Li2TiO3 with Excess Lithium

• Li 2 TiO 3 with excess Li was heated at 900 °C for 30 days in 1000 Pa H 2 /Ar flow. • Li mass loss was 0.7 wt%, grain size increased and BET surface area decreased. • The majority of evaporative Li was lost by first 3 days. • The influence of the 30 days heating on tritium behavior was not large. • Tritium recovery by isotope exchange with H 2 was effective at 600 °C and 900 °C. Tritium breeding ceramic materials are placed at high temperatures for a long period in a fusion DEMO reactor. Therefore, the understanding of Li mass loss phenomena and its influence on tritium behavior are important. In this study, the pebbles of Li 2 TiO 3 with excess Li were heated at 900 °C for 30 days in a 1000 Pa H 2 /Ar flow and tritium sorption and recovery experiments were carried out. Li mass loss by the heating was evaluated to be 0.7 wt%. The value of Li mass loss was almost same as that for 3 days heating at the same condition. Tritium sorption capacity for the heated pebbles at 600 °C and 900 °C were almost same as that for the pebbles as received. Tritium sorbed in the pebbles could not be recovered effectively by the 1000 Pa H 2 /Ar purge at room temperature and 300 °C but it could be recovered at 600 °C and 900 °C. The influence of the long-time heating on the behavior of tritium sorbed in the pebbles was not large.

Activation analysis of the European DEMO divertor with respect to the different breeding blanket segmentation

• Neutronic analysis of European DEMO fusion reactor single-null type divertor cassette. • Divertor neutron activation calculation with three different integrated breeding blanket modules (HCPB, WCLL MMS, WCLL SMS). • Neutron activation in divertor functional layers (radionuclide and reaction analysis). Neutron activation is an unavoidable process in large scale fusion devices operating with deuterium and tritium fuels. Neutron activation leads to the production of radioactive materials. Activated materials can generate heat and ionizing radiation in their environment. Consequently, values of activation characteristics need to be determined in order to ensure safety and performance of the fusion devices. In this work activity and decay heat inventories for the DEMOnstration power plant (DEMO) 2015 baseline model divertor were calculated. Operation scenario of 1898 days was considered as well as three different breeder blanket configurations of DEMO reactor: single-module segmentation water cooled lithium lead (WCLL SMS), multi-module segmentation water cooled lithium lead (WCLL MMS) and helium cooled pebble bed (HCPB). Divertor model is divided into 62 segments and 4 layers with different material compositions. Neutron transport and activation inventory calculations were performed respectively with MCNP and FISPACT codes. In addition to activation calculation results, divertor radionuclide analysis is also presented in this work.

Magneto-convective effect on tritium transport at breeder unit level for the WCLL breeding blanket of DEMO

The Water-Cooled Lithium-Lead (WCLL) is one of the four breeding blanket concepts proposed by Europe in view of its DEMO reactor. The velocity field of the electrically conducting lead-lithium eutectic alloy inside the blanket is strongly influenced by the external magnetic field used for plasma confinement combined with buoyancy effect. The strength of the magnetohydrodynamics (MHD) effect and of the magneto-convective effect (MHD and buoyancy force) depends on the intensity of the magnetic field and its orientation with respect to the direction of the lead-lithium motion. This phenomenon significantly influences the resulting temperature and velocity fields, and therefore the tritium transport inside the breeding blanket. A multi-physics approach of a 3D tritium transport model is presented for a simplified geometry of the WCLL breeding blanket. In particular, advection-diffusion of tritium into the lead-lithium eutectic alloy, transfer of tritium from the liquid interface towards the steel, diffusion of tritium inside the steel, transfer of tritium from the steel towards the coolant, and advection-diffusion of diatomic tritium into the coolant, temperature field, velocity fields of both lead-lithium and water, buoyancy forces, and MHD effect have been included in this study. The tritium concentrations and the inventories inside the lead-lithium, in the Eurofer pipes and in the baffle, and in the water coolant have been evaluated.

Thermomechanical characterisation of W-Eurofer 97 brazed joints

The preliminary conceptual design of both the blanket and the divertor of the future DEMO reactor envisages tungsten as plasma-facing material joined to Eurofer 97 steel as the structural element. Therefore, the development of new joining technologies to constitute these first wall components has become an essential task in the fusion community. In this regard, there is a lack of knowledge on the thermal stability and mechanical behaviour of these W-Eurofer 97 joints under the expected service temperatures, which will be situated around 400–600 °C. In this paper, successful diffusion brazing of Eurofer 97 steel with tungsten was achieved by using a copper interlayer. Its microstructure, hardness and flexural strength were investigated for different configurations and thermal histories in its predicted operating temperature range.Investigations on as-brazed specimens showed defect-free interfaces and a continuous metallic bond between W and Eurofer 97 was successfully accomplished; although the hardening of steel base material was detected after joining. Microstructure alterations were observed just at the W-Eurofer 97 interface, with the diffusion of Cu braze into the iron grain boundaries and the formation of a Fe-rich hard intermetallic layer, which was, in fact, the weakest part of the joint during thermo-mechanical testing.The flexural strength of the brazed specimens at room temperature was relatively low, with a strong dependence on the anisotropy of the W base material and little influence of the stress-relieving thermal treatment. However, testing at temperatures up to 600 °C showed high strength of the bonded specimens, with values as high as 470 MPa, comparable to those of the Eurofer 97 steel.

Mechanical properties of several newly produced RAFM steels with Tungsten content in the range of 2 wt%

Abstract The contribution of ENEA together with Rina-CSM to the Eurofusion programme “WPMAT-Advanced Seels” deals with the development of innovative RAFM steels able to withstand the critical temperatures typical of the different operational environments foreseen for the blanket of the first DEMO reactor. The optimization of the chemical composition and the Thermo Mechanical Treatment for these materials should be done according to the blanket operating temperatures that are related to two possible working conditions: the WCLL-BB (Water Cooled Lead Lithium Breeding Blanket) or the H(D)CLL-BB (Helium (Dual) Cooled Lead Lithium Breeding Blanket). On the one hand the “water-cooling” option implies a minimum irradiation temperature for the blanket material in the range of 280–350 °C. On the other hand, the “helium-cooled” and the “dual-coolant” solutions imply an operating temperature for the blanket material in the range of 650 °C. Therefore in the first case the target is the improvement of the toughness of the martensitic alloys; whilst concerning the second scenario the target is the development of more creep resistant martensitic steels, suitable to tolerate such a high operating temperature. In both the cases the Tungsten content plays a key role, both in terms of solid solution hardening and influence on the DBTT. Two alloys aimed at fulfilling the specifications for the two DEMO operating conditions, both with increased Tungsten content respect to Eurofer, have been produced and characterized. The mechanical properties of these two alloys are hereby reported and discussed.

Influence of radiation defects on deuterium permeation behavior in tungsten

Understanding the hydrogen isotope (H) permeation behavior in neutron irradiated tungsten (W) is highly desired for the accurate prediction of H behavior in W first walls in a fusion DEMO reactor. Contrary studies have been published where some indicated a relatively higher H permeability in the irradiated W whereas others showed a lower H permeability. In these studies the radiation defects were produced with ions (as surrogates for neutrons) and were concentrated in a shallow layer (<200 nm), their effects could be mixed up with surface effects during H permeation. Herein W specimens damaged to 800 nm by Au-ion implantation are adopted to investigate the effects of radiation defects on deuterium (D) permeation. Microstructural analyses with slow positron annihilation technique and transmission electron microscopy revealed that vacancy-type defects and dislocation loops were produced in the implanted W. Results from D2 gas-driven permeation measurements at 700–1000 K showed, surprisingly, radiation defects did not influence the D permeability. The phenomenon was illustrated from a thermodynamic perspective. Radiation-induced dislocations and/or monovacancies, possessing low H coverage fractions, do not affect permeability because their effects on effective diffusivity and solubility can fully cancel each other. Vacancy clusters, having a high fractional H coverage and owing to their strong trapping nature, will not come to effect during the steady-state permeation portion, thus do not contribute to permeability.

Analysis of the influence of the different impurity seeding on the burn-up fraction and plasma confinement in the EU DEMO reactor

Abstract This work describes integrated numerical modelling applied to DEMO discharges with tungsten wall and divertor, using the COREDIV code, which self-consistently solves 1D radial transport equations of plasma and impurities in the core region and 2D multi-fluid transport in the SOL. The model is self-consistent with respect to both the effects of impurities on the α-power level and the interaction between seeded (Ar, Kr and Xe) and intrinsic impurities (tungsten, helium). This work is to analyse the influence of the impurity seeding on the burn-up fraction and plasma confinement in EU DEMO reactor. For the simulation with constant electron density at the separatrix, it is found that impurity seeding has a small influence on the burn-up fraction, which remains around 6 – 7%., but fuelling source is reduced from 2.7 to 2 × 1022 1/sec when moving from lowest to highest seeding level. The degradation of confinement for high Z seeding impurity, which is correlated to radiation in the core (due to seeded Kr and Xe) is observed. Better confinement for Ar at middle and high seeding level is observed in comparison to the case without seeding, because Ar radiation is small in the core (about 40-55 MW).

Parametric study of the influence of double-walled tubes layout on the DEMO WCLL breeding blanket thermal performances

Abstract Within the framework of the EUROfusion activities regarding the EU-DEMO Breeding Blanket (BB) concept, the University of Palermo is long-time involved, in close cooperation with ENEA, in the design of the Water Cooled Lithium Lead (WCLL) BB, that is one of the two concepts under consideration for the DEMO reactor. It is mainly characterized by a liquid lithium-lead eutectic alloy acting as breeder and neutron multiplier, as well as by subcooled pressurized water flowing as coolant under PWR-like conditions (pressure of 15.5 MPa and inlet/outlet temperatures of 295 °C/328 °C). A research campaign has been recently carried out to study the potential influence of the Breeding Zone cooling circuit layout, composed of bundles of Double Walled Tubes (DWTs), on the blanket thermo-mechanical behaviour. In the first phase, a parametric study has been performed to investigate the effects of some geometrical DWT parameters onto the blanket thermal response in order to identify the best tubes layout, in case arranged along toroidal-radial planes. In the second phase, the thermo-mechanical performances of the blanket equatorial region, equipped with the optimized DWTs layout, have been assessed. The study has been carried out following a theoretical-numerical approach based on the Finite Element Method and adopting the ABAQUS v. 6.14 commercial code. Model, assumptions and results obtained are herewith presented and critically discussed.

Preliminary development of a conceptual first wall for DEMO

The first wall is one of the key components in a tokamak, in particular, for DEMO in the future. The conceptual first wall for a DEMO reactor is designed, fabricated and tested in this paper. It has a sandwich-like structure, which consists of a reduced activation ferritic/martensitic (RAFM) substrate, an interlayer and a plasma-facing material of tungsten produced by chemical vapor deposition (CVD-W) due to its characteristics of high density, high purity and superior thermal shock resistance. The interlayer is required to not only have good adhesion between CVD-W and the CLF-1 substrate, which is a type of RAFM developed by the Southwestern Institute of Physics, but also the capability of tritium transport prevention due to the great importance of controlling the tritium buildup in the first wall, improving the fuel efficiency and conforming to the safety regulations of tritium. Titanium nitride (TiN), a well-known tritium barrier, is selected as the interlayer material in this work. In total, eight sandwich-like first wall samples have been fabricated, and thermal cycle tests and plasma exposure tests are implemented. The results of material analysis show that the quality of the TiN coating layer is critical for good adhesion and tritium prevention. At the defects of the TiN interlayer, fluorine (F) penetrated into the CLF-1 during the CVD-W fabrication and, as a result, the sandwich-like structure was broken near the CLF-1 side. For some samples, the TiN interlayer made by CVD has sufficient adhesion as an interlayer between CVD-W and CLF-1 during tests so far. These encouraging test results support further examination of TiN as an interlayer material for W on an RAFM steel as a first wall material. Preliminary results indicate that the design and fabrication of the conceptual first wall are feasible. Further tests, including the improvement of the interlayer quality and the tritium permeation test, are necessary in the future.

Estimation of magnetic error field with alleviating fabrication tolerance of large superconducting magnets on JA DEMO reactor

Abstract Generally, DEMO requires larger toroidal field (TF) coils than ITER, resulting in one of the major difficulties, the tolerance in TF coil fabrication. This paper presents the possible solutions based on the design study on Japan’s DEMO (JA DEMO). It was confirmed that, in the case of adopting a mitigated tolerance by a factor of 2.5–5 compared with that of ITER, the resulting error field of TF coils is correctable to an acceptable level in terms of locked mode avoidance. In addition, the design of the error field correction coil (EFCC) on JA DEMO was investigated.