Tobita, Kenji; Federici, G.*; Okano, Kunihiko
Fusion Engineering and Design, 89(9-10), p.1870 - 1874, 2014/10
The goal of the DEMO reactor design under the Broader Approach (BA) is to develop possible pre-conceptual designs of DEMO by addressing key design issues and options in physics, technology and system engineering for DEMO. The joint work between EU and Japan for the DEMO design started with a benchmark of systems codes. Cross-checking between the EU systems code PROCESS and the JA systems code TPC showed a good agreement for relatively conservative plasma parameters. In parallel, critical design issues on DEMO have been studied. In order to resolve the problem on divertor heat removal, a reduction of divertor heat load due to plasma detachment and advanced divertor concepts such as super-X and snowflake configuration has been investigated. Regarding remote maintenance (RM), various RM concepts based on different sector segmentations and access ports has been studied to allow reasonable plant availability under severe in-vessel dose rate.
Nakamura, Makoto; Tobita, Kenji; Gulden, W.*; Watanabe, Kazuhito*; Someya, Yoji; Tanigawa, Hisashi; Sakamoto, Yoshiteru; Araki, Takao*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.
Fusion Engineering and Design, 89(9-10), p.2028 - 2032, 2014/10
After the Fukushima Dai-ichi nuclear accident, a social need for assuring safety of fusion energy has grown gradually in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of BA 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 amounts of radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO, in which the blanket technology is based on the Japanese fusion technology R&D programme. Reference event sequences expected in DEMO have been analyzed based on the master logic diagram and functional FMEA techniques. Accident initiators of particular importance in DEMO have been selected based on the event sequence analysis.
Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Sakamoto, Yoshiteru; Araki, Takao*; Watanabe, Kazuhito*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.
Plasma and Fusion Research (Internet), 9, p.1405139_1 - 1405139_11, 2014/10
Key aspects of the safety study of a water-cooled fusion DEMO reactor is reported. Safety requirements, dose target, DEMO plant model and confinement strategy of the safety study are briefly introduced. The internal hazard of a water-cooled DEMO, i.e. radioactive inventories, stored energies that can mobilize these inventories and accident initiators and scenarios, are evaluated. It is pointed out that the enthalpy in the first wall/blanket cooling loops, the decay heat and the energy potentially released by the Be-steam chemical reaction are of special concern for the water-cooled DEMO. An ex-vessel loss-of-coolant of the first wall/blanket cooling loop is also quantitatively analyzed. The integrity of the building against the ex-VV LOCA is discussed.
Nakamura, Makoto; Kemp, R.*; Uto, Hiroyasu; Ward, D. J.*; Tobita, Kenji; Hiwatari, Ryoji*; Federici, G.*
Fusion Engineering and Design, 87(5-6), p.864 - 867, 2012/08
For fusion research directed at electricity generation in the ITER and post-ITER era, it is necessary to define development targets toward DEMO including plasma parameters and engineering requirements such as magnetic field and divertor heat flux. In general as a first step of systematic reactor design, systems analysis is performed in order to estimate reactor operation windows with engineering constraints. Thus, evaluation of existing systems analysis codes or development of systems codes is essential for basis of fusion reactor plasma parameters and engineering requirements. In this paper we report recent our efforts towards improvement of systems codes for the BA DEMO design, i.e. benchmarking the two systems codes the Japan and EU home teams are managing. The main result is that calculation outputs from the two codes are in good agreement.
Ioki, Kimihiro*; Barabash, V.*; Cordier, J.*; Enoeda, Mikio; Federici, G.*; Kim, B. C.*; Mazul, I.*; Merola, M.*; Morimoto, Masaaki*; Nakahira, Masataka*; et al.
Fusion Engineering and Design, 83(7-9), p.787 - 794, 2008/12
This paper presents recent results of ITER activities on Vacuum Vessel (VV), blanket, limiter, and divertor. Major results can be summarized as follows. (1) The VV design is being developed in more details considering manufacturing and assembly methods, and cost. Incorporating manufacturing studies being performed in cooperation with parties, the regular VV sector design has been nearly finalized. (2) The procurement allocation of blanket modules among 6 parties was fixed and the blanket module design has progressed in cooperation with parties. Fabrication of mock-ups for prequalification testing is under way and the tests will be performed in 2007-2008. (3) The divertor activities have progressed with the aim of launching the procurement according to the ITER project schedule.
Kamiya, Kensaku; Asakura, Nobuyuki; Boedo, J. A.*; Eich, T.*; Federici, G.*; Fenstermacher, M.*; Finken, K.*; Herrmann, A.*; Terry, J.*; Kirk, A.*; et al.
Plasma Physics and Controlled Fusion, 49(7), p.s43 - s62, 2007/07
Edge Localized Mode (ELM) measurements in the tokamaks, including JT-60U, DIII-D, ASDEX-U and JET, are reviewed. The followings are outlines of this presentation. (1) ELM Types and basic scaling, (2) Small ELM regimes and ELM mitigation, (3) ELM filament formation and transverse motion, (4) Power deposition on divertor targets and main chamber wall.
Loarte, A.*; Lipschultz, B.*; Kukushkin, A. S.*; Matthews, G. F.*; Stangeby, P. C.*; Asakura, Nobuyuki; Counsell, G. F.*; Federici, G.*; Kallenbach, A.*; Krieger, K.*; et al.
Nuclear Fusion, 47(6), p.S203 - S263, 2007/06
Progress, since the ITER Physics Basis publication (1999), in understanding the processes that will determine the properties of the plasma edge and its interaction with material elements in ITER is described. Significant progress in experiment area: energy and particle transport, the interaction of plasmas with the main chamber material elements, ELM energy deposition on material elements and the transport mechanism, the physics of plasma detachment and neutral dynamics, the erosion of low and high Z materials, their transport to the core plasma and their migration at the plasma edge, retention of tritium in fusion devices and removal methods. This progress has been accompanied by the development of modelling tools for the physical processes at the edge plasma and plasma-materials interaction. The implications for the expected performance in ITER and the lifetime of the plasma facing materials are discussed.
Lipschultz, B.*; Asakura, Nobuyuki; Bonnin, X.*; Coster, D. P.*; Counsell, G.*; Doerner, R.*; Dux, R.*; Federici, G.*; Fenstermacher, M. E.*; Fundamenski, W.*; et al.
Proceedings of 21st IAEA Fusion Energy Conference (FEC 2006) (CD-ROM), 8 Pages, 2007/03
The work of the ITPA SOL/divertor group is reviewed. The high-n nature of ELMs has been elucidated and new measurements have determined that they carry 10-20% of the ELM energy to the far SOL with implications for ITER limiters and the upper divertor. Analysis of ELM measurements imply that the ELM continuously loses energy as it travels across the SOL. The prediction of ITER divertor disruption power loads have been reduced as a result of finding that the divertor footprint broadens during the thermal quench and that the plasma can lose up to 80% of its thermal energy before the thermal quench (not for VDEs or ITBs). Disruption mitigation through massive gas puffing has been successful at reducing divertor heat loads but estimates of the effect on the main chamber walls indicate 10s of kG of Be would be melted/mitigation. Long-pulse studies have shown that the fraction of injected gas that can be recovered after a discharge decreases with discharge length. The use of mixed materials gives rise to a number of potential processes.
Iida, Hiromasa; Petrizzi, L.*; Khripunov, V.*; Federici, G.*; Polunovskiy, E.*
Fusion Engineering and Design, 75(1-4), p.133 - 139, 2005/11
The design of the ITER machine was presented in 2001. A nuclear analysis has been performed on ITER by means of the most detailed models and the best assessed nuclear data and codes. As the construction phase of ITER is approaching, the design of the main components has been optimized/finalized and several minor design changes/optimizations have been made, which required refined calculations to confirm that nuclear design requirements are met. Some of the proposed design changes have been made to mitigate critical radiation shielding problems. This paper reviews some of the most recent neutronic work with emphasis on critical nuclear responses in the TF coil inboard legs and vacuum vessel related to design modifications made to the blanket modules and vacuum vessel.
Shimada, Michiya; Costley, A. E.*; Federici, G.*; Ioki, Kimihiro*; Kukushkin, A. S.*; Mukhovatov, V.*; Polevoi, A. R.*; Sugihara, Masayoshi
Journal of Nuclear Materials, 337-339, p.808 - 815, 2005/03
ITER is an experimental fusion reactor for investigation and demonstration of burning plasmas, characterised of its heating dominated by alpha-particle heating. ITER is a major step from present devices and an indispensable step for fusion reactor development. ITER's success largely depends on the control of plasma-wall interactions(PWI), with power and particle fluxes and time scales one or two orders of magnitude larger than in present devices. The strategy for control of PWI includes the semi-closed divertor, strong fuelling and pumping, disruption and ELM control, replaceable plasma-facing materials and stepwise operation.
Roth, J.*; Kirschner, A.*; Bohmeyer, W.*; Brezinsek, S.*; Cambe, A.*; Casarotto, E.*; Doerner, R.*; Gauthier, E.*; Federici, G.*; Higashijima, Satoru; et al.
Journal of Nuclear Materials, 337-339, p.970 - 974, 2005/03
In the frame work of the EU Task Force on Plasma-Wall Interaction and the International Tokamak Physics Activity an attempt was made to establish a possible dependence of the chemical erosion yield of carbon on the ion flux, , involving ion beam experiments, plasma simulators, and fusion devices. After data normalization a fit using Bayesian probability analysis was performed yielding a decrease of the erosion yield with at high ion fluxes. With this dependence on ion flux a comprehensive description is available for chemical erosion as function of energy, temperature and ion flux. Using this dependence the erosion and redeposition of carbon in the ITER divertor can be calculated using the ERO code and the steady-state plasma scenario given by the ITER team. The resulting gross and net erosion rates are compared to previous estimates using a constant erosion yield of 1.5%. The use of the complete parameter dependence results in an order of magnitude lower erosion, most strongly determined by the temperature dependence and the reduction at the highest fluxes.
Roth, J.*; Preuss, R.*; Bohmeyer, W.*; Brezinsek, S.*; Cambe, A.*; Casarotto, E.*; Doerner, R.*; Gauthier, E.*; Federici, G.*; Higashijima, Satoru; et al.
Nuclear Fusion, 44(11), p.L21 - L25, 2004/11
Chemical erosion of carbon has been studied in ion beam experiments, and the yield values are available as a function of ion energy and surface temperature. ITER divertor condition, however, cannot be simulated by ion beam. For extrapolating to ITER, the erosion must be investigated in plasma simulators and in SOL or divertors of present fusion devices. In the past, erosion values were reported, but the values showed a wide scatter as a function of ion flux, . Therefore, a joint attempt was made through the EU Task Force on Plasma-Wall Interaction and the International Tokamak Physics Activity (ITPA) to clarify the flux dependence. For each data point the local plasma conditions were normalized to impact energy of 30 eV, the data were selected for a surface temperature close to the maximum yield or to room temperature, and the diagnostic was calibrated in-situ. Through this procedure, the previous large scatter could be drastically reduced. A fit using Bayesian probability analysis was performed yielding a decrease of the erosion yield with at high ion fluxes.
Shimada, Michiya; Campbell, D.*; Stambaugh, R.*; Polevoi, A. R.*; Mukhovatov, V.*; Asakura, Nobuyuki; Costley, A. E.*; Donn, A. J. H.*; Doyle, E. J.*; Federici, G.*; et al.
Proceedings of 20th IAEA Fusion Energy Conference (FEC 2004) (CD-ROM), 8 Pages, 2004/11
This paper summarises recent progress in the physics basis and its impact on the expected performance of ITER. Significant progress has been made in many outstanding issues and in the development of hybrid and steady state operation scenarios, leading to increased confidence of achieving ITER's goals. Experiments show that tailoring the current profile can improve confinement over the standard H-mode and allow an increase in beta up to the no-wall limit at safety factors 4. Extrapolation to ITER suggests that at the reduced plasma current of 12MA, high Q 10 and long pulse (1000 s) operation is possible with benign ELMs. Analysis of disruption scenarios has been performed based on guidelines on current quench rates and halo currents, derived from the experimental database. With conservative assumptions, estimated electromagnetic forces on the in-vessel components are below the design target values, confirming the robustness of the ITER design against disruption forces.
Ioki, Kimihiro*; Akiba, Masato; Barabaschi, P.*; Barabash, V.*; Chiocchio, S.*; Daenner, W.*; Elio, F.*; Enoeda, Mikio; Ezato, Koichiro; Federici, G.*; et al.
Journal of Nuclear Materials, 329-333(1), p.31 - 38, 2004/08
The preparation of the procurement specifications is being progressed for key components. Progress has been made in the preparation of the procurement specifications for key nuclear components of ITER. Detailed design of the vacuum vessel (VV) and in-vessel components is being performed to consider fabrication methods and non-destructive tests (NDT). R&D activities are being carried out on vacuum vessel UT inspection with waves launched at an angle of 20 or 30 degree, on flow distribution tests of a two-channel model, on fabrication and testing of FW mockups and panels, on the blanket flexible support as a complete system including the housing, on the blanket co-axial pipe connection with guard vacuum for leak detection, and on divertor vertical target prototypes. The results give confidence in the validity of the design and identify possibilities of attractive alternate fabrication methods.
Shimada, Michiya; Mukhovatov, V.*; Federici, G.*; Gribov, Y.*; Kukushkin, A.*; Murakami, Yoshiki*; Polevoi, A. R.*; Pustovitov, V. D.*; Sengoku, Seio; Sugihara, Masayoshi
Nuclear Fusion, 44(2), p.350 - 356, 2004/02
Recent performance analysis has improved confidence in achieving Q 10 in inductive operation in ITER. Performance analysis based on empirical scaling shows the feasibility of achieving Q 10 in inductive operation with a sufficient margin. Theory-based core modeling indicates the need of high pedestal temperature (2-4 keV) to achieve Q 10, which is in the range of projection with pedestal scaling. The heat load of type-I ELM could be made tolerable by high density operation and further tilting the target plate (if necessary). Pellet injection from High-Field Side would be useful in enhancing Q and reducing ELM heat load. Steady state operation scenarios have been developed with modest requirement on confinement improvement and beta (HH98(y,2) 1.3 and betaN 2.6). Stabilisation of RWM, required in such regimes, is feasible with the present saddle coils and power supplies with double-wall structure taken into account.
Shimada, Michiya; Mukhovatov, V.*; Federici, G.*; Gribov, Y.*; Kukushkin, A. S.*; Murakami, Yoshiki*; Polevoi, A. R.*; Pustovitov, V. D.*; Sengoku, Seio; Sugihara, Masayoshi
Nuclear Fusion, 44(2), p.350 - 356, 2004/02
Performance analysis based on empirical scaling shows the feasibility of achieving Q 10 in inductive operation. Analysis has also elucidated a possibility that ITER can potentially demonstrate Q 50, enabling studies of self-heated plasmas. Theory-based core modeling indicates the need of high pedestal temperature (3.2 - 5.3 keV) to achieve Q 10, which is in the range of projection with presently available pedestal scalings. Pellet injection from high-field side would be useful in enhancing Q and reducing ELM heat load in high plasma current operation. If the ELM heat load is not acceptable, it could be made tolerable by further tilting the target plate. Steady state operation scenarios at Q = 5 have been developed with modest requirement on confinement improvement and beta (HH98(y,2) 1.3 and betaN 2.6). Stabilisation of RWM, required in such regimes, is feasible with the present saddle coils and power supplies with double-wall structure taken into account.
Mukhovatov, V.*; Shimada, Michiya; Chudnovskiy, A. N.*; Costley, A. E.*; Gribov, Y.*; Federici, G.*; Kardaun, O. J. F.*; Kukushkin, A. S.*; Polevoi, A. R.*; Pustovitov, V. D.*; et al.
Plasma Physics and Controlled Fusion, 45(12), p.235 - 252, 2003/12
ITER will be the first magnetic confinement device with burning DT plasma and fusion power of about 0.5 GW. During the past few years, new results have been obtained that substantiate the confidence in achieving Q 10 in ITER with inductive H-mode operation. These include achievement of a good H-mode confinement near the Greenwald density at high triangularity of the plasma cross section; improvements in theory-based confinement projections for the core plasma; improvement in helium ash removal due to the elastic collisions of He atoms with D/T ions in the divertor predicted by modelling; demonstration of feedback control of NTMs and resultant improvement in the achievable beta-values; better understanding of ELM physics and development of ELM mitigation techniques; and demonstration of mitigation of plasma disruptions. ITER will have a flexibility to operate also in steady state and intermediate (hybrid) regimes. The paper concentrates on inductively driven plasma performance and discusses requirements for steady-state operation in ITER.
Loarte, A.*; Saibene, G.*; Sartori, R.*; Campbell, D.*; Becoulet, M.*; Horton, L.*; Eich, T.*; Herrmann, A.*; Matthews, G.*; Asakura, Nobuyuki; et al.
Plasma Physics and Controlled Fusion, 45(9), p.1549 - 1569, 2003/10
Analysis of Type I ELMs from ongoing experiments shows that ELM energy losses are correlated with the density and temperature of the pedestal plasma before the ELM crash. The Type I ELM plasma energy loss normalized to the pedestal energy is found to correlate across experiments with the collisionality of the pedestal plasma. Other parameters affect the ELM size such as the edge magnetic shear, etc, which influence the plasma volume affected by the ELMs. ELM particle losses are influenced by this ELM affected volume and are weakly dependent on other pedestal plasma parameters. In JET and DIII-D, minimum Type I ELMs with energy losses acceptable for ITER were found, that do not affect the plasma temperature. The duration of the divertor ELM power pulse is correlated with the typical ion transport time from the pedestal to the divertor target and not with the duration of the ELM associated MHD activity. Extrapolation of the present experimental results to ITER is summarized.
Loarte, A.*; Saibene, G.*; Sartori, R.*; Becoulet, M.*; Horton, L.*; Eich, T.*; Herrmann, A.*; Laux, M.*; Matthews, G.*; Jachmich, S.*; et al.
Journal of Nuclear Materials, 313-316, p.962 - 966, 2003/03
no abstracts in English
Asakura, Nobuyuki; Loarte, A.*; Porter, G.*; Philipps, V.*; Lipschultz, B.*; Kallenbach, A.*; Matthews, G.*; Federici, G.*; Kukushkin, A.*; Mahdavi, A.*; et al.
IAEA-CN-94/CT/P-01, 5 Pages, 2002/00
Three important physics issues for the ITER divertor design and operation are summarized based on the experimental and numerical work from multi-machine database (JET, JT-60U, ASDEX Upgrade, DIII-D, Alcator C-Mod and TEXTOR). (i) The energy load associated with Type-I ELMs is of great concern for the lifetime of the ITER divertor target. In order to understand the physics base of the scaling models, the ELM heat and particle transport to the divertor is investigated. Convective transport during ELMs plays an important role in heat transport to the divertor. (ii) Determination of the SOL flow pattern and the driving mechanism has progressed experimentally and numerically. Influences of the drift effects on the SOL and divertor plasma transport were discussed. (iii) Characteristics of chemical yield at two different deposited carbon surfaces, i.e. erosion- and redeposition-dominated areas, have been studied. Progress of understanding the chemical erosion is reviewed.