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The current status and the progress of research and development (R&D) activities for a Fusion DEMO reactor in the National Institutes for Quantum and Radiological Science and Technology (QST) Rokka...

Requirements for pellet injection system for a Japanese DEMO concept are comprehensively assessed in order to elucidate direction and target of R&D. A fusion output control simulation study using 1.5D transport code is carried out to estimate requirement of fuel particle source. Next, multivariable parameter dependence survey of the fuel deposition source on pellet speed, poloidal injection angle, pellet mass and pedestal height is performed. The conditions to achieve the target fueling depth derived from the output control study are shown. The AUG calibrated relation is referred to consider spatial restriction of tokamak geometry which relates maximum pellet velocity to curvature of pellet guide tube. Then the restriction is collated to the results of multivariable survey to find the most advantageous pellet speed, injection angle, and pellet mass which is consistent with both output control and tokamak geometry. Consequently, a conceptual design of fuel pellet injection system as the R&D target toward to DEMO is proposed.

This paper reports the current status of a safety research on water-cooled fusion DEMO in Japan. A basic strategy of development of the safety guidelines is described for DEMO based on a water-cooled solid pebble bed blanket. Clarification of safety features of the DEMO in accident situations is a key issue to develop the guidelines. Recent achievements in understanding of the safety features of the water-cooled DEMO are reported. The MELCOR analysis has clarified the temperature histories of the DEMO components in a complete loss of decay heat removal event. The transient behavior of the first wall temperature is found to be essentially different from that of ITER. The pressure load to the tokamak cooling water system vault (TCWSV) is analyzed based on a simple model equation of the energy conservation. If the amount of the primary coolant is the same as that of Slim-CS, the previous small Japanese DEMO, the discharged water does not damage the TCWSV with the volume and pressure-tightness similar to those of pressurized light water reactors. It is shown that implementation of a pressure suppression system to the small TCWSV is effective to suppress the pressure load to the second confinement barrier.

Plasma position control for DEMO reactor has been investigated using numerical simulation, which consists of plasma equilibrium, eddy current and active feedback control analyses. The stabilization effect of in-vessel components, the influence on the magnetic detector and the power of active feedback control coils are evaluated. Especially, the influence of breeding blanket modules on plasma position control is shown in this paper.

In order to realize a feasible DEMO, we designed an in-vessel component including the conducting shell. The project is affiliated with the broader approach DEMO design activities and is conceptualized from a plasma vertical stability and engineering viewpoint. The dependence of the plasma vertical stability on the conducing shell parameters and the electromagnetic force at plasma disruption were investigated in numerical simulations (programmed in the 3D eddy current analysis code and a plasma position control code). The simulations assumed the actual shape and position of the vacuum vessel and in-vessel components. The plasma vertical stability was most effectively maintained by the double-loop shell.

The conceptual design of Japan’s fusion demonstration plant (JA DEMO) is now being developed. In this paper, an overall plant system concept related to tritium handling in the water-cooling system is developed to give a concrete shape to the present JA DEMO concept as an electric power plant. The basic condition of tritium permeation from the in-vessel components to the primary cooling system is evaluated to be 5.7 g-T/day. The tritium concentration of the primary coolant is assumed to be 1 TBq/kg similar to the heavy water reactor condition. The capacity of the water detritiation system (WDS) is assessed, and the bypass feed water from the primary cooling loop is evaluated to be 94 kg/h under the tritium extraction efficiency of 0.96. Based on those specific parameters, the existing WDS in the heavy water reactor is found to be applicable to that of JA DEMO. Configuration of the primary heat transfer system (PHTS) is also discussed. Based on the heavy water reactor experience, tritium permeation through a steam generator (SG) to the secondary cooling system in PHTS is evaluated at 11.77TBq/year/loop (318 Ci/year/loop), which is found to be less than the restricted amount of tritium disposal for a pressurized water reactor in Japan. The key effect of the heavy water reactor experience is reduction of tritium permeation by oxide layer formed on SG pipes. Finally, a confinement concept of tritium release from PHTS is discussed under the condition of ex-vessel loss of coolant accident (LOCA). A pressure suppression system is installed to prevent the upper tokamak hall from pressurizing at the ex-vessel LOCA, and the tritium leakage from the upper tokamak hall is consequently restrained. The resultant early public dose at the plant site boundary can be reduced to 1.8 mSv, which is negligibly smaller than 100 mSv of the no-evacuation limit recommended by IAEA.

Blanket concept with a simplified interior for mass production has been developed using a mixed bed of Li2TiO3 and Be12Ti pebbles, coolant conditions of 15.5 MPa and 290–325 °C and cooling pipes without any partitions. Considering the continuity with the ITER test blanket module option of Japan and the engineering feasibility in its fabrication, our design study focused on a water-cooled solid breeding blanket using the mixed pebbles bed. Herein, we propose blanket segmentation corresponding to the shape and dimension of the blanket and routing of the coolant flow. Moreover, we estimate the overall tritium breeding ratio (TBR) with a torus configuration, based on the segmentation using three-dimensional (3D) Monte Carlo N-particle calculations. As a result, the overall TBR is 1.15. Our 3D neutronics analysis for TBR ensures that the blanket concept can achieve a self-sufficient supply of tritium.

Maintenance schemes are one of the critical issues in DEMO design, significantly affecting the configuration of in-vessel components, the size of toroidal field (TF) coil, the arrangement of poloidal field (PF) coils, reactor building, hot cell and so forth. Therefore, the maintenance schemes should satisfy many design requirements and criteria to assure reliable and safe plant operation and to attain reasonable plant availability. The plant availability depends on reliability of remote maintenance scheme, inspection of pipe connection and plasma operation. In this paper, various remote maintenance schemes for DEMO were comparatively assessed based on requirements for DEMO remote maintenance. From the view points of the reliability of inspection on hot cell, TF coil size, stored energy of PF coil and portability of segment, the banana shape segment transport using all vertical maintenance ports would be more probable DEMO reactor maintenance scheme, and it has key engineering issues such as in-vessel transferring mechanism of segment, pipe connection and conducting shell design for plasma vertical stability.

We have considered a strategy for reducing the radioactive waste generated by the replacement of in-vessel components, such as blanket segments and divertor cassettes, for the fusion DEMO reactor. ...

A short super-X divertor (SXD) is proposed as an option for the Demo divertor, where the field line length from the divertor null to the outer target was largely increased compared to a similar-size conventional divertor. Physics and engineering design studies for a 3 GW-level fusion power Demo reactor (SlimCS) (Tobita et al., 2009) have recently progressed. Minimal number of the divertor coils were installed inside the toroidal field coil, i.e. interlink-winding. Arrangement of the poloidal field coils and their currents were determined, taking into account of the engineering design such as vacuum vessel and the neutron shield structures, and the divertor maintenance scenario. Divertor plasma simulation showed that significant radiation region is produced between the super-X null and the target. Radiation loss in the divertor was increased, producing fully detached plasmas efficiently. Advantages of the short SXD were demonstrated, but the total peak heat load was a marginal level (10 MW m −2 ) for the engineering design.

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.

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 (Li 2 TiO 3 ) and beryllium–titanium (Be 12 Ti) 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.

Management strategy of radioactive waste generated in periodic replacement may be important in the point of view of fusion reactor design, because it has a large impact on the design of the hot cell and waste storage in the plant. In the replacement period of a fusion power reactor, the assembly of blanket or divertor modules needs to be removed from the reactor in order to minimize remote maintenance in the vacuum vessel and to attain reasonable plant availability. In the hot cell, the modules will be removed from the back plate of the assembly. Here, note that the active cooling must be done by a way that does not cause contamination of the hot cell environment due to dispersion of tritium and tungsten dust. In this sense, the cooling scenario is adopted that the existing pipe of cooling water in the assembly is connected to a different cooling water system in the hot cell. On the other hand, it is assumed that the structural material (F82H) of the blanket and divertor is not recycled due to its high contact dose rate. It should be crushed into small pieces to reduce volume of the waste and required storage space. In this paper, the basic idea of the waste management scenario and the conceptual design in the hot cell and waste storage for DEMO has been proposed.

To determine the engineering feasibility of an “inter-linked” (IL) superconducting central solenoid (CS) and poloidal field (PF) coil, we studied the basic engineering design of the IL-CS and PF coil. We were able to use IL superconducting coils in a tokamak fusion demonstration power plant (DEMO) reactor to achieve a sufficient amount of the CS magnetic flux swing for the current ramp-up while maintaining plasma size and also achieving an advanced divertor concepts of “super-X divertor (SXD)” by locating near the plasma. A basic idea of the IL superconducting CS and PF coil concept is to wind a CS (PF coil) such that it is linked in a set of toroidal field (TF) coils. In this study, we considered the conductor and feeder design of the IL-CS and PF coil including the superconducting strand type, cooling, and winding and assembly methods, and determined that react and wind Nb3Al is a candidate for use in an IL coil and feeder in this specification. This paper presents the preliminary engineering design results of the IL superconducting CS and PF coil for creating a SXD equilibrium configuration.

Relations between a net electrical output power and dimensions of components in radial build are investigated based on the ITER plasma performance to develop a conceptual design of DEMO with the net electrical output power of several hundred MW. Reducing the dimensions of in-vessel components and increasing the thickness of the toroidal field coil contribute to strengthen the toroidal magnetic field at plasma, which brings about increase in a net electrical output power. The relation between the minimum plasma major radius and the maximum net electrical output power is clarified. Furthermore effects of improvements in the ITER plasma performance on the net electricity are also analyzed; indicating the increase of normalized beta could have advantage from the viewpoint of the divertor heat load because the increase of synchrotron radiation loss power contributes to reduce the divertor heat load, though the higher energy confinement is required.

Design study of the magnetic configuration and divertor geometry for the “advanced divertor” in a Demo tokamak reactor is summarized. Equilibrium calculation code, TOSCA, was developed for the supe...

The maintenance scheme is a critical issue for DEMO design, and requires high availability of the reactor. The SlimCS, designed in JAEA, adopts the horizontal sector transport hot cell maintenance scheme. In order to determine the most appropriate DEMO reactor maintenance scheme, it is important to assess the various maintenance schemes. In this paper the vertical sector transport maintenance concept is proposed for the first time. In the sector maintenance scheme, the amount of cutting/re-welding of the piping is minimized. The sector including blanket modules and a high-temperature shield was divided into 10° segments in a toroidal direction. The sectors are designed to be removed and re-inserted through upper alternate vertical maintenance ports. In the case of the vertical sector transport maintenance scheme, inter-coil structures could be adopted for use against turnover force in toroidal field (TF) coils. This shows an advantage in DEMO maintenance compared with horizontal sector transport.

Waste is generated at the moment when the operation of a fusion reactor is halted and maintenance is started for periodic replacement of blanket modules and divertor. Used blanket and divertor need to be replaced shortly after the shutdown for high plant availability, as long as high surface dose rate and decay heat of the blanket and divertor can be handled. In this sense, nuclear characteristics of the blanket and divertor need to be understood for a reasonable maintenance scheme. For the purpose, neutronic calculations were carried out on the blanket and divertor using a THIDA-2 code with FENDL-2.0. For a SlimCS DEMO reactor, the calculated decay heat for each 1/12-sector was as high as 5 MW just after the shutdown and 0.3 MW one month later. For the maintenance, a cooled shielding structure (CSS) was proposed to remove the decay heat and to shield gamma-rays from the sector. When maintenance is done one month after the shutdown, the sector temperature is maintained to be 550 °C or lower with the cooling by the CSS of 50 °C. In order to avoid tritium release from the sector during the maintenance, a cask should be used to transport the sector. For efficient use of resources, breeding and neutron multiplying materials should be reused or recycled. A possible strategy for reuse or recycle is also presented.

For the water-cooled solid blanket of DEMO, the nuclear analysis was performed based on present cooling piping system. Especially, distributions of neutron load and temperature were calculated with Pn is 5 MW/m 2 . Furthermore, the local TBR was optimized by changing the material proportion for each Pn level (1–5 MW/m2). It was confirmed that the size of cooling loop for sub-critical water could be used as about 2000 × 450 mm and the cooling pipe diameter of D is 12 mm, d is 9 mm at v is 5.36 m/s. The pipe pitches would vary with Pn level which is related to the blanket structure design. Nuclear heat distribution is the base to decide the distribution of cooling pipe positions. It was found that the local TBR of blanket would be dropped down along with the Pn level rising which was mainly depended on the thickness of beryllium variation. Finally, the layout of cooling pipes for each level was obtained.

Conceptual design of an innovatively small tokamak reactor “GNOME” based on a non-fission biomass-fusion hybrid concept is proposed. This fusion plant concept intends to use high-temperature heat from the blanket to generate hydrogen or synthetic fuels out of waste biomass. Since energy multiplication is expected by utilizing chemical energy of biomass, the requirement for the fusion plasma for net plant energy output is reduced to Q ≥ 5. As a result, the GNOME reactor has been designed to produce 320 MW fusion power with a 5.2 m major radius, 3.1 normalized beta and 11 T maximum field. This relatively small maximum field can be achieved by using Nb3Sn superconducting magnets. Besides, this reactor allows 3.0 m diameter space for its center solenoid coil and requires 60 MW of the input power. These features require minimal technical extensions from ITER.

Simplification of blanket system is necessary for a fusion DEMO reactor. Although the conceptual design of a tritium-breeding blanket for SlimCS has been studied in the past several years, the structure of the previous blanket seems to be complex and difficult to manufacture from the viewpoint of engineering. In this paper, we proposed simplification of blanket structure without decreasing the net Tritium Breeding Ratio (TBR). In the proposed concept, the blanket is filled with the mixture of Li 4 SiO 4 pebbles or Li 2 O pebbles for the tritium breeding and Be 12 Ti pebbles for the neutron multiplication. To confirm the effectiveness of this concept, an ANIHEAT code with the nuclear library FENDL-2.0 was used for calculations of the neutronic and thermal analyses. The result indicated that, under the constraint of the blanket thickness being less than 0.5 m, the mixture of Li 2 O pebbles and Be 12 Ti ones is the most effective and that the TBR is expected to be greater than 1.05.

For a tokamak fusion DEMO reactor with the fusion output of 2.95 GW, neutronic and thermal design of blanket is under way to find a feasible blanket concept. For the continuity with the Japanese ITER-TBM options, this study considered water-cooled blanket with solid breeding materials of Li ceramics (Li 4 SiO 4 , Li 2 TiO 3 and Li 2 ZrO 3 ) and Be multipliers (Be and Be 12 Ti). Based on a neutronics-heat coupled analysis, the tritium breeding ratio was evaluated so as to satisfy constraints of the operating temperature of ≤900 °C for Li ceramics and Be 12 Ti, and ≤600 °C for Be. Cooling water condition was assumed to be 23 MPa and 290-360 C. The result indicates that surplus tritium production in lower neutron wall load (P n ) blanket compensates a shortfall in higher P n blanket and thus the overall tritium production can marginally satisfy fuel self-sufficiency. .

a b s t r a c t Nuclear analysis results were compared for water-cooled blanket based on PWR (pressurized water reactor) and SCWR (sub-critical water reactor) water conditions. The local TBR (tritium breeding ratio) in outboard zone was discussed in the range of Pn (neutron wall load) from 1 MW/m 2 to 5 MW/m 2 . It was found that water fraction has little impact on TBR, which is an important factor related to blanket tritium efficiency. It indicated that TBR value of each Pn would be similar under the two kinds of water conditions, but PWR case is a little higher than that of SCWR’s. In addition, it was found that beryllium is the dominant factor leading a higher TBR inside blanket. As a result, TBR is an insensitive value with the water condition variation. The results would be important to water condition choice for solid blanket in the future.

This paper summarizes the evolution of Japanese DEMO design studies in a retrospective manner by highlighting efforts to resolve critical design issues on DEMO. Japan is currently working on the conceptual study of a steady-state DEMO (JA DEMO) with a major radius Rp of 8.5 m and fusion power Pfus of 1.5 to 2 GW based on water-cooled solid breeding blanket with pressurized water reactor water condition (290ºC to 325ºC, 15.5 MPa). Such a lower Pfus allows to find realistic design solutions for divertor heat removal. Recognizing that divertor heat removal is one of the most challenging issues on DEMO, the divertor design has been carried out in different approaches, including numerical divertor plasma simulation, magnetic configurations, heat sink design, etc. It is noteworthy that the latest divertor simulation led to a design window allowing divertor heat removal of the peak heat flux of

In this contribution, the work done in AINA code during 2014 and 2015 at IFERC is presented. The main motivation of this work was to adapt the code and to perform safety studies for a Japanese DEMO design. Related to AINA code, the work has supposed major changes in plasma models. Significant is the addition of an integrated SOL-pedestal model that allows the estimation of heat loads at divertor. Also, a thermal model for a WCPB (water cooled pebble bed) breeding blanket has been developed based in parametric input data from neutronics calculations. Related to safety studies, a major breakthrough in the study of LOPC (loss of plasma control) transients has been the use of an optimization method to determine the most severe transients in terms of the shortest melting times. The results of the safety study show that LOPC transients are not likely to be severe for breeding blanket, but for the case of divertor can induce severe melting. For ex-vessel LOCA (loss of coolant accident) analysis, it is severe for both blanket and divertor, but in the first case the transient time until melting is nearly two orders of magnitude higher. The results point out that the recovery time for plasma control system should be at least one order of magnitude lower than confinement time to avoid melting of divertor targets.