Development and Application of SONIC Divertor Simulation Code to Power Exhaust Design of Japanese DEMO Divertor

An integrated divertor simulation code, SONIC, has been developed in order to predict a self-consistent transport solution of the plasma, neutral and impurities in the scrape-off layer (SOL) and divertor. SONIC code has contributed to determining the divertor design and power handling scenarios for the Japanese (JA) fusion demonstration (DEMO) reactor. Radiative cooling scenario of Ar impurity seeding and the divertor performance have been demonstrated to evaluate the power exhaust scenarios with Psep = 230–290 MW. The simulation identified the decay length of the total parallel heat flux profile as being broader than the electron one, because of the ion convective transport from the outer divertor to the upstream SOL, produced by the plasma flow reversal. The flow reversal also reduced the impurity retention in the outer divertor, which may produce the partial detachment. Divertor operation margin of key power exhaust parameters to satisfy the peak qtarget ≤ 10 MWm−2 was determined in the low nesep of 2 − 3 × 1019 m−3 under severe conditions such as reducing radiation loss fraction, i.e., f*raddiv = (Pradsol + Praddiv)/Psep and diffusion coefficients (χ and D). The divertor geometry and reference parameters (f*raddiv ~ 0.8, χ = 1 m2s−1, D = 0.3 m2s−1) were consistent with the low nesep operation of the JA DEMO concepts. For either severe assumption of f*raddiv ~ 0.7 or χ and D to their half values, higher nesep operation was required. In addition, recent investigations of physics models (temperature-gradient force on impurity, photon transport, neutral–neutral collision) under the DEMO relevant SOL and divertor condition are presented

Assessment on Tokamak Fusion Power Plant to Contribute to Global Climate Stabilization in the Framework of Paris Agreement

A cost model for a tokamak fusion power plant (FPP) is improved to evaluate material cost and manufacture cost, separately. Then, the improved cost model is applied to a commercial tokamak FPP, and reduction of FPP construction cost is investigated considering learning effect on manufacture of the fusion island part and the advanced manufacture of toroidal field (TF) coils. Finally, a development scenario of a tokamak FPP is proposed to contribute substantially to global climate stabilization under the framework of the Paris Agreement

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

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 (Rp/ap= 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 Psep~283MW) and JA-DEMO with higher plasma elongation (a revised design with Psep~235MW). The divertor operation with the peak qtarget < 10MWm-2 was determined in the low nesep of 2-3x1019 m-3 under the severe conditions of reducing radiation loss fraction, i.e. f*raddiv = (Pradsol+Praddiv)/Psep, and diffusion coefficients (chi and D). The divertor geometry and reference key parameters (f*raddiv ~0.8, chi =1 m2/s and D = 0.3 m2/s) were so far consistent with the power exhaust concepts in the nesep range, and the revised JA-DEMO design has advantages of wider nesep range and enough margin for the divertor operation. For either severe assumption of f*raddiv~0.7 or chi and D to the half value, higher nesep operation was required for the primary design in order to control the peak qtarget < 10MWm-2, i.e. the operation window was reduced. Applying the two severe assumptions, the divertor operation was difficult in the low nesep range

Plasma Scenario Modeling for JA DEMO

In order to proceed the integrated plasma scenario modeling of JA DEMO, (i) core transport simulation, (ii) vertical stability and (iii) MHD stability analyses have been performed. In addition, applicability of no/small ELM regimes to JA DEMO has also been investigated.On the 1.5-D time-dependent core transport simulation by integrated code TOPICS, the steady-state operation condition with HH = 1.41, betaN = 3.6, fBS = 0.69 is demonstrated by optimizing the heating scenario, where CDBM transport model is used. It is also demonstrated that off-axis ECCD has important roles for maintaining the internal transport barriers (ITBs) for steady-state condition and for controlling the fusion power by control of ITB location. On the vertical stability analysis, the ramp-up scenario of high elongated plasma has been developed by using the plasma equilibrium simulator MECS with 3D eddy current effects. The temporal evolutions of the poloidal beta and internal inductance are evaluated using TOPICS. The result indicates that plasma elongation at 95% of poloidal flux of 1.75 is achievable in JA DEMO. Regarding the MHD stability analysis, the beta limit of JA DEMO plasma has been evaluated by using the linear ideal MHD stability code MARG2D. The beta limit without conducting wall is betaN ~ 2.6, while that with conducting wall can be improved to ~3.5 at the wall radius of rW/a = 1.35. Further improvements are observed with decreasing the wall radius, for example betaN ~ 3.9 at rW/a = 1.30.It is important to investigate the applicability of no/small ELM regimes to a DEMO. The candidates of the no/small ELM regimes realized in the low collisionality region are grassy ELMy H mode and QH mode, which are observed in many tokamak experiments. The grassy ELM regime can be reduced the ELM energy to ~1% of the pedestal energy. The necessary conditions are the high poloidal beta ~1.4, the high safety factor q95>4, the high triangularity >0.4 and the low elongation <1.45. The DEMO main parameters, that satisfy the necessary conditions of grassy ELMy H mode, will be discussed.6th IAEA DEMO Programme Worksho

Plasma Scenario Modeling for JA DEMO

In order to proceed the integrated plasma scenario modeling of JA DEMO, (i) core transport simulation, (ii) vertical stability and (iii) MHD stability analyses have been performed. In addition, applicability of no/small ELM regimes to JA DEMO has also been investigated.On the 1.5-D time-dependent core transport simulation by integrated code TOPICS, the steady-state operation condition with HH = 1.41, betaN = 3.6, fBS = 0.69 is demonstrated by optimizing the heating scenario, where CDBM transport model is used. It is also demonstrated that off-axis ECCD has important roles for maintaining the internal transport barriers (ITBs) for steady-state condition and for controlling the fusion power by control of ITB location. On the vertical stability analysis, the ramp-up scenario of high elongated plasma has been developed by using the plasma equilibrium simulator MECS with 3D eddy current effects. The temporal evolutions of the poloidal beta and internal inductance are evaluated using TOPICS. The result indicates that plasma elongation at 95% of poloidal flux of 1.75 is achievable in JA DEMO. Regarding the MHD stability analysis, the beta limit of JA DEMO plasma has been evaluated by using the linear ideal MHD stability code MARG2D. The beta limit without conducting wall is betaN ~ 2.6, while that with conducting wall can be improved to ~3.5 at the wall radius of rW/a = 1.35. Further improvements are observed with decreasing the wall radius, for example betaN ~ 3.9 at rW/a = 1.30.It is important to investigate the applicability of no/small ELM regimes to a DEMO. The candidates of the no/small ELM regimes realized in the low collisionality region are grassy ELMy H mode and QH mode, which are observed in many tokamak experiments. The grassy ELM regime can be reduced the ELM energy to ~1% of the pedestal energy. The necessary conditions are the high poloidal beta ~1.4, the high safety factor q95>4, the high triangularity >0.4 and the low elongation <1.45. The DEMO main parameters, that satisfy the necessary conditions of grassy ELMy H mode, will be discussed.6th IAEA DEMO Programme Worksho

Simulation Study of the Divertor Operation for a DEMO Fusion Reactor

Power handling in the divertor has been investigated for Japan DEMO concept. System code suggests the plasma line-av. ne (8.6x1019m-3~1.2nGW) lower than ITER. Plasma detachment in the low nesep (ne/3) = 2-3x1019m-3 has been investigated. Appropriate divertor size, i.e. larger than ITER-size leg, has been investigated. Divertor size will affect sizes of vacuum vessel, toroidal field coil and maintenance port as well as its cost. Divertor performance with leg length (Ldiv) of 1.6m was investigated in low SOL density (nesep = 1.5-2.4x10m-3), using SONIC with Ar seeding, Pout = 250MW (Psep= 235-240MW), high radiation fraction of 0.8 and fixed diffusion coefficients. High plasma temperature at SOL (Tesep = 300-600eV): local decay length of the parallel heat flux profile is 2.6-2.8 mm near the separatrix due to electron flux limiter, which is smaller than ITER simulation (3.6 mm, Tesep ~200 eV). Peak heat load at the dovertor target was reduced from 12 to ~5 MWm-2 with increasing separatrix density (nesep). nesep >2.2x1019m-3 was required to reduce Tediv to ~25 eV at the attached region of the outer target, and to reduce nAr/ne in SOL to ~0.5%. “Detachment plasma (~1eV)” was produced near the separatrix for both Ldiv = 1.1m and 1.6m cases, while higher Te and Ti at attached plasma area for the shorter case. Ar concentration in SOL was larger (~3%) for the short Ldiv = 1.1m case.The 23rd International Conference on Plasma Surface Interactions in Controlled Fusion Devices (PSI-23

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

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

Development and Application of SONIC Divertor Simulation Code to Power Exhaust Design of Japanese DEMO Divertor

An integrated divertor simulation code, SONIC, has been developed in order to predict a self-consistent transport solution of the plasma, neutral and impurities in the scrape-off layer (SOL) and divertor. SONIC code has contributed to determining the divertor design and power handling scenarios for the Japanese (JA) fusion demonstration (DEMO) reactor. Radiative cooling scenario of Ar impurity seeding and the divertor performance have been demonstrated to evalu-ate the power exhaust scenarios with Psep = 230–290 MW. The simulation identified the decay length of the total parallel heat flux profile as being broader than the electron one, because of the ion convective transport from the outer divertor to the upstream SOL, produced by the plasma flow reversal. The flow reversal also reduced the impurity retention in the outer divertor, which may produce the partial detachment. Divertor operation margin of key power exhaust parameters to satisfy the peak qtarget less than 10 MWm−2 was determined in the low nesep of 2 − 3 × 1019 m−3 under severe conditions such as reducing radiation loss fraction, i.e., f*raddiv = (Pradsol + Praddiv)/Psep and diffu-sion coefficients (chi and D). The divertor geometry and reference parameters (f*raddiv ~ 0.8, chi = 1 m2s−1, D = 0.3 m2s−1) were consistent with the low nesep operation of the JA DEMO concepts. For either severe assumption of f*raddiv ~ 0.7 or chi and D to their half values, higher nesep operation was required. In addition, recent investigations of physics models (temperature-gradient force on impurity, photon transport, neutral–neutral collision) under the DEMO relevant SOL and divertor condition are presented

Cooling water system design of Japan\u27s DEMO for fusion power production

Considering the water cooling system in fusion reactors, there are several fusion-specific challenges in designs of in-vessel components and their cooling water systems (CWS). In this research, solutions of the challenges have been discussed and indicate the CWS concept of Japan\u27s DEMO. The CWS transfers thermal energy of blanket and a part of divertor to a generator system, and thermal energy of the other part of divertor is used as heat utilization in plant. The required performance of system devices such as pumps is expected to be achieved by proven technologies and its extensions. It is also indicated that blanket can work as the low-pass filter, and can suppress the effect of fusion output fluctuation. The designed cooling water system requires 70 MW electric power and generates 620 MW power with a 1.5 GW fusion power plasma. This research indicates the basic concept of Japan\u27s DEMO CWS that can be practically achieved