DEMO Fusion Reactor Research Articles

Attachment system for helium-cooled blanket of RF DEMO fusion reactor

The development of DEMO thermonuclear reactor is a part of Russian national program on the fusion process mastering. The DEMO-S (stationary thermonuclear reactor) should be the logic continuation of the ITER-type projects (pulse thermonuclear reactors) and the prototype for commercial power plants. DEMO reactor layout suggests to use the segmented blanket with mounting/dismounting procedure through the vacuum vessel vertical ports. Taking into account this layout the blanket attachment system has been developed and the present paper is devoted to this subject. The considered attachment system includes the lower and upper toroidal support assemblies which connect all the blanket segments in the enclosed ring. In it's turn the lower support assemblies attached to the vacuum vessel through the system of hinged support pillars. The heights of support pillars for inboard and outboard blankets are selected so that to indemnify the blanket massif thermal expansions in vertical and radial directions. The support pillars have been calculated on strength taking into account the electromagnetic loads from the plasma disruptions and blanket mass. The selection of high-strength chromium steel as a structural material for the support pillars could be considered as the results of strength analysis. The conclusions on the possibility to apply this attachment system for fusion reactor blanket and the critical issues are contained in this paper too.

Design and R&D issues for the JT-60 modification to a full superconducting tokamak

This paper describes the modification program of JT-60 towards the tokamak fusion DEMO-reactor in support of ITER. The JT-60 is planned to be modified for the superconducting tokamak, JT-60SC, to establish steady-state operation at high beta and ensure plasma applicability of ferritic steel as reduced activation materials in a reactor-relevant regime with a break-even class plasma. Main parameters of the JT-60SC device are the plasma current of 4 MA, the toroidal field of 3.8 T, the major and minor radii of 2.8 and 0.85 m, respectively, the elongation and triangularity of 1.8 and 0.35, respectively, a current flattop duration of 100 s and the total heating power of 15 MW for 100 s. Nb 3Al and Nb 3Sn superconducting strands with a high Cu/non-Cu ratio and NbTi strands with ∼10 μm diameter of filaments have been developed for the toroidal and poloidal superconducting coils requiring compactness and low AC losses. Full-scale cable-in-conduit cables were manufactured by automatic welding of SS316 conduits and jacketing the cables. An irradiation test for a mockup of the cost-effectively designed CFC divertor target plate has demonstrated the required high heat handling capability of ∼10 MW/m 2 on the JAERI Electron Beam Irradiation Stand.

Challenges for insulating materials in fusion applications

It is envisaged that early in this new century International Thermonuclear Experimental Reactor (ITER) will come into operation to bridge the gap between the present day large “physics” machines and the pre-commercial DEMO fusion reactor. Although ITER will undoubtedly help to solve many of the problems which still remain in the field of plasma physics, it will also present additional operational and experimental problems due to radiation damage effects as a result of the intense radiation field from the “burning” plasma. For structural metallic materials the problem of radiation damage is expected to be severe, although tolerable, only near to the first wall, however the problem facing the numerous insulating components is far more serious due to the necessity to maintain not only the mechanical, but also the far more sensitive physical properties intact. Insulating materials will be required in a number of key systems ranging from heating and current drive, to diagnostics, and remote handling which affect not only the operation, but also the safety and control of the machine, as well as maintenance and repair. The effects of radiation on the electrical and optical properties will be discussed, in particular radiation induced conductivity (RIC), radiation induced electrical degradation (RIED), radiation induced electromotive force (RIEMF) and radioluminescence.

Hydrogen permeation through tritium permeation barrier in Pb–17Li

One of the main problems in the development of water cooled lithium lead (WCLL) DEMO fusion reactor is the reduction of the tritium permeation from the Pb–17Li, or the plasma, into the cooling water. The control of tritium losses is an important issue in fusion technology because of its safety and operational implications. This goal can be achieved using a tritium permeation barrier (TPB). The use of aluminium rich coatings, which forms Al 2O 3 at their surface, has been selected as reference solution for WCLL blanket in order to produce reliable TPB. The hot dipping process is one of the two candidates for the production of coatings on large blanket segments. The effectiveness of hot dipped aluminium coating on MANET II steel has been verified in gas phase and in liquid Pb–17Li, using a test apparatus named Corelli II. The permeation rate measured in gas phase is one order of magnitude lower than the one in liquid metal phase. Performing SEM-EDS analysis on the specimen, it was observed that micro cracks on the coating surface were present. The permeation curves in gas and Pb–17Li are reported and discussed. The characteristics of the new experimental device Vivaldi will be briefly described.

Nuclear and Thermal Analyses of Supercritical-water-cooled Solid Breeder Blanket for Fusion DEMO Reactor

Within a design study of a fusion DEMO reactor aiming at demonstrating technologies of fusion power plant, supercritical water is applied as a coolant of solid breeder blanket to attain high thermal efficiency. The blanket has multi-layer composed of solid breeder pebbles (Li2O) and neutron multiplier pebbles (Be) which are radially separated by cooling panels. The first wall and the breeding region are cooled by supercritical water below and above the pseudo- critical temperature, respectively. Temperature distribution and tritium breeding ratio (TBR) have been estimated by one-dimensional nuclear and thermal calculations. The local TBR as high as 1.47 has been obtained after optimization of temperature distribution in the breeder region under the following conditions: neutron wall loading of 5 MW/m2, 6Li enrichment of 30% and coolant temperature at inlet of breeder region of 380°C. In the case of the higher coolant temperature 430°C of the breeder region the local TBR was reduced to be 1.40. This means that the net TBR higher than 1.0 could be expected with the supercritical-water-cooled blanket, whose temperature distribution in the breeder region would be optimized by following the coolant temperature, and where a coverage of the breeder region is assumed to be 70%.

High dose neutron irradiation damage in beryllium as blanket material

The paper presents the investigation results of beryllium products that operated in the SM and BOR-60 reactors up to neutron doses of 2.8×10 22 and 8.0×10 22 cm −2 ( E>1 MeV), respectively. The calculated and experimental data are given on helium and tritium accumulation, swelling, micro-hardness and thermal conductivity. The microstructural investigation results of irradiated beryllium are also presented. It is shown that the rate of helium and tritium accumulation in beryllium in the SM and BOR-60 reactors is high enough, which is of interest from the viewpoint of modeling the working conditions of the DEMO fusion reactor. Swelling of beryllium at irradiation temperature of 70–150 °C and neutron fluence of 2.8×10 22 cm −2 ( E>1 MeV) makes up 0.8–1.5%, at 400 °C and fluence of 8×10 22 cm −2 ( E>1 MeV) – 3.2–5.0%. Irradiation hardening and decrease of thermal conductivity strongly depend on the irradiation temperature and are more significant at reduced temperatures. All results presented in the paper were analyzed with due account of the supposed working parameters of the DEMO fusion reactor blanket.

DEMO blanket testing in ITER. Influence on reaching DEMO

ITER goal was specified as one step between now and the DEMO fusion reactor. One of the major issues is the tritium breeding blankets test relevant to future reactors. The major objectives of blanket modules (TBM) experiments in ITER are reduced in comparison with proposed test objectives in ITER-FDR. Thus, results of DEMO blanket designs testing in ITER will provide limited (but still useful) information that will need strong support from non-fusion facilities testing. The role of non-fusion tests is increased now to provide additional data required for DEMO blanket construction and qualification. A strategy of testing steps to DEMO blanket qualifications has to include parallel testing in ITER and in non-fusion devices. Experiments in fission reactors are able to provide essential data on materials radiation properties; tritium release, inventory and permeation; and thermomechanical behavior of the blanket breeder/multiplier. However, the volume in fission reactors is rather small and neutron spectra differ from the fusion reactor one. Nonetheless in the near future one depends primarily on fission reactor irradiation. The powerful accelerator based neutron source IFMIF could also provide useful information on radiation material properties. Plasma based neutron sources of different fusion devices could be the best choice for testing DEMO materials and blanket mock-ups. Timetable and costs of these devices are not clear now.

Nuclear and Thermal Analyses of Supercritical-water-cooled Solid Breeder Blanket for Fusion DEMO Reactor.

Within a design study of a fusion DEMO reactor aiming at demonstrating technologies of fusion power plant, supercritical water is applied as a coolant of solid breeder blanket to attain high thermal efficiency. The blanket has multi-layer composed of solid breeder pebbles (Li2O) and neutron multiplier pebbles (Be) which are radially separated by cooling panels. The first wall and the breeding region are cooled by supercritical water below and above the pseudo- critical temperature, respectively. Temperature distribution and tritium breeding ratio (TBR) have been estimated by one-dimensional nuclear and thermal calculations. The local TBR as high as 1.47 has been obtained after optimization of temperature distribution in the breeder region under the following conditions: neutron wall loading of 5 MW/m2, 6Li enrichment of 30% and coolant temperature at inlet of breeder region of 380°C. In the case of the higher coolant temperature 430°C of the breeder region the local TBR was reduced to be 1.40. This means that the net TBR higher than 1.0 could be expected with the supercritical-water-cooled blanket, whose temperature distribution in the breeder region would be optimized by following the coolant temperature, and where a coverage of the breeder region is assumed to be 70%.

High heat flux test of a HIP-bonded first wall panel of reduced activation ferritic steel F-82H

Reduced activation ferritic steel F-82H is a primary candidate structural material of DEMO fusion reactors. In fabrication technology, development of the DEMO blanket in JAERI, a hot isostatic pressing (HIP) bonding method, especially for the first wall structure with built-in cooling tubes has been proposed. A HIP-bonded F-82H first wall panel was successfully fabricated with selected manufacturing parameters. A high heat flux test of the HIP-bonded F-82H first wall panel has been performed to examine the thermo-mechanical performance of the panel including the integrity of the HIP-bonded interfaces and the fatigue behavior. A maximum heat flux of 2.7 MW/m2 was applied to accelerate the fatigue test up to 5000 cycles in test blanket inserted ITER. The maximum temperature of the panel was ∼450°C under this heat flux. Through this test campaign, no damage such as cracks was observed on the surface of the panel, and no degradation in heat removal performance was observed either from the temperature responses. The thermal fatigue lifetime of the panel was found to be longer than the fatigue data obtained by mechanical testing.

Progress in the design of the HCPB test blanket module mock-ups for the HE-FUS3 facility

One of the ENEA contributions in the activity planned for the European Fusion Technology Programme for 1997–1998 is the design and testing of helium cooled pebble bed blanket (HCPB) mock-ups for the DEMO fusion reactor. The reference medium scale mock-up, called HEXCALIBER (HE-FUS3 experimental cassette of lithium beryllium pebble beds), will reproduce a portion of the HCPB module with the L 4SiO 4 ceramic breeder and the beryllium neutron multiplier, both in the form of pebble beds heated by flat heaters. The feasibility of the final design and the manufacturing solutions will be first evaluated by testing a reduced portion of the mock-up called HELICA (HE-FUS3 lithium cassette) filled with L 4SiO 4 breeder pebbles. Aimed at defining the thermal map of the mock-up and the related stresses, some preliminary FEM theoretical calculations were carried-out by ANSYS code. This paper presents the design of two mock-ups, the preliminary calculations, the raw material behaviours and the qualification of the manufacturing procedures. The experimental tests will be carried-out on the HE-FUS3 facility at ENEA Brasimone.

Development of tritium permeation barriers on Al base in Europe

The development of the water cooled lithium lead (WCLL) DEMO fusion reactor requires the production of a material capable of acting as a tritium permeation barrier (TPB). In the DEMO blanket reactor permeation barriers on the structural material are required to reduce the tritium permeation from the Pb–17Li or the plasma into the cooling water to acceptable levels (<1 g/d). Because of experimental work previously performed, one of the most promising TPB candidates is A1 base coatings. Within the EU a large R&D programme is in progress to develop a TPB fabrication technique, compatible with the structural materials requirements and capable of producing coatings with acceptable performances. The research is focused on chemical vapour deposition (CVD), hot dipping, hot isostatic pressing (HIP) technology and spray (this one developed also for repair) deposition techniques. The final goal is to select a reference technique to be used in the blanket of the DEMO reactor and in the ITER test module fabrication. The activities performed in four European laboratories are summarised here.

Users' requirements for IFMIF

The International Fusion Materials Irradiation Facility (IFMIF) is a high energy neutron irradiation facility which generates an intense neutron flux with D–Li stripping reactions for fusion materials testing. The role of IFMIF is (1) development of various fusion reactor materials, (2) determination of design-relevant engineering databases for the DEMO fusion reactor, (3) calibration and validation of data generated from fission reactor irradiations and the other simulation experiments, etc. The conceptual design activity (CDA) of IFMIF was initiated in February 1995 as an IEA collaborative activity to complete a reference conceptual design of IFMIF in December 1996. Users' requirements for the conceptual design of IFMIF were developed for materials to be tested, types of experiments, small specimen test technology and irradiation conditions. Furthermore, the neutron irradiation field characteristics (spectrum, flux/volume, etc.) of IFMIF were evaluated for the conceptual design parameters and were shown to meet the essential requirements of the users.

Current status and future R&D for reduced-activation ferritic/martensitic steels

International research and development programs on reduced-activation ferritic/martensitic steels, the primary candidate-alloys for a DEMO fusion reactor and beyond, are briefly summarized, along with some information on conventional steels. An International Energy Agency (IEA) collaborative test program to determine the feasibility of reduced-activation ferritic/martensitic steels for fusion is in progress and will be completed within this century. Baseline properties including typical irradiation behavior for Fe–(7–9)%Cr reduced-activation ferritic steels are shown. Most of the data are for a heat of modified F82H steel, purchased for the IEA program. Experimental plans to explore possible problems and solutions for fusion devices using ferromagnetic materials are introduced. The preliminary results show that it should be possible to use a ferromagnetic vacuum vessel in tokamak devices.

European helium cooled pebble bed blanket: Design of blanket: module to be tested in ITER

The helium-cooled pebble bed (HCPB) blanket for a fusion DEMO reactor is based on the use of separate small lithium orthosilicate and beryllium pebble beds placed between radial toroidal cooling plates. The cooling is provided by helium at 8 MPa. The tritium produced in the pebble beds is purged by the flow of helium at 0.1 MPa. The structural material is martensitic steel. It is foreseen, after extensive R&D, to test in ITER a blanket module based on the HCPB design which is one of the two European proposals for the ITER Test Blanket Programme. To facilitate the handling operations, the blanket test module (BTM) is bolted to a surrounding water cooled frame fixed to the ITER shield blanket back plate. For the design of the test module, 3D Monte Carlo neutronic calculations and thermo-hydraulic and stress analyses for the operation during the basic performance phase (BPP) and during the extended performance phase (EPP) of ITER have been performed. The behaviour of the test modules during LOCA, LOFA and electromagnetic transients has been investigated. Conceptual designs of the required ancillary loops have been performed.

Stresses in the First Wall of a Dual-Coolant Liquid-Metal Breeder Blanket during Electron-Beam Welding

Electron-beam (EB) welding is an important joining technique for fabricating and assembling blanket and first-wall components in fusion reactors. A numerical modeling of the EB procedure of a selected part of a DEMO fusion reactor blanket is presented. Stress and temperature distributions during and after EB welding are analyzed with the help of nonlinear finite element calculations, including phase transformations of MANET stainless steel. Residual stresses are discussed. Their magnitude and distribution may stimulate the initiation and growth of surface cracks parallel to the weld. Analysis of postweld heat treatment shows that the residual stresses can be neglected if appropriate postweld heat treatment is performed. The main drawback of this analysis seems to be the lack of appropriate material data for high temperatures close to the melting point. Despite this, qualitative statements on the feasibility of joints are possible, and improved analyses of welding stresses are expected to make a valuable contribution to future studies that address the safety and reliability of blanket structures.

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.

Demo & Commercial Fusion Reactors Extrapolated from the ITER and Advanced Physics & Materials Data Bases

Demo & Commercial Fusion Reactors Extrapolated from the ITER and Advanced Physics & Materials Data Bases

Status of EC solid breeder-blanket designs and R&amp;D for DEMO fusion reactors

Status of EC solid breeder-blanket designs and R&amp;D for DEMO fusion reactors

Status of EC solid breeder-blanket designs and R&amp;D for DEMO fusion reactors

Within the European Community Fusion Technology Programme, two solid breeder-blankets for a DEMO reactor are being developed. The two blankets have various features in common: helium as coolant and as tritium-purge gas; the martensitic steel MANET as a structural material; and beryllium as the neutron multiplier. The configurations of the two blankets, however, are different: in the breeder inside tube concept, the breeder materials are LiA1O 2 or Li 2 ZrO 3 in the form of annular pellets contained in tubes surrounded by beryllium blocks, with the coolant helium being outside the tubes; in contrast, in the breeder out of tube concept, the breeder and multiplier materials are Li 4 SiO 4 and beryllium pebbles, forming a mixed bed placed outside the tubes containing the coolant helium. The main critical issues for both blankets are the behaviour of the breeder ceramics, and of beryllium under irradiation and the tritium control. Other issues are the low temperature irradiation-induced embrittlement of the MANET; the mechanical effects caused by major plasma disruptions; and safety and reliability. The R&D work concentrates on these issues. The development of martensitic steels, including MANET, is part of a separate programme. Breeder ceramics and beryllium irradiations have been performed so far under conditions which do not cover the peak values projected for the DEMO blankets. Further irradiations in thermal reactors and in fast reactors, especially for beryllium, are required. Effective tritium control requires the development of permeation barriers and/or of methods of oxidation of the tritium in the main helium cooling system. First promising results have been obtained also in the field of mechanical effects from plasma disruptions and safety and reliability. However, further work is required in the reliability field and to validate the codes for the calculations of the plasma disruption effects.

Remote handling technology development for fusion experimental reactors

Fusion experimental reactors are a major step toward realization of fusion energy production by commerical reactors. The key objective is to demonstrate the scientific and technological feasibility prior to the demo fusion reactors. The ITER (international thermonuclear experimental reactor) is an experimental reactor and the engineering design has been initiated by the united efforts of the USA, EC, RF, and Japan after completion of the conceptual design activity. Since the reactor core components are activated by the D–T operation, advanced remote handling technologies are inevitably required for assembly and maintenance of the components under the radiation conditions. The key issues to be developed for the remote handling system are radiation hardening over 10 7 R h −1, precise position—posture control for handling heavy components without the interference in a limited reactor space, and tools for connection—disconnection and inspection. For this purpose, extensive technology developments in parallel with the engineering design are being conducted domestically and internationally through the ITER framework. This paper outlines both an overview of the remote handling system required for fusion experimental reactors and the recent status of remote handling technology development being conducted in Japan, EC and USA.