Power Exhaust Concepts and Divertor Designs for Japanese and European DEMO Fusion Reactors

Concepts of the power exhaust and divertor design have been developed, with a high priority in the pre-conceptual design phase of the Japan-Europe Broader Approach DEMO Design Activity. A common critical issue is the large power exhaust and its fraction in the main plasma and divertor by the radiative cooling. Different exhaust concepts in the main plasma and divertor have been developed for JA and EU DEMOs. JA proposed a conventional closed divertor geometry to challenge large Psep/Rp handling of 30-35 MWm-1 in order to maintain the radiation fraction in the main plasma at the ITER-level (fradmain = Pradmain/Pheat ~0.4) and higher plasma performance. EU challenged both increasing fradmain to ~0.65 and handling the ITER-level Psep/Rp in the open divertor geometry. Power exhaust simulations have been performed by SONIC (JA) and SOLPS5.1 (EU) with corresponding Psep = 250-300 MW and 150-200 MW, respectively. Both results showed that large divertor radiation fraction (Praddiv/Psep 0.8) was required to reduce both peak qtarget ( 10MWm-2) and Te,idiv. In addition, the JA divertor performance with EU-reference Psep of 150MW showed benefit of the closed geometry to reduce the peak qtarget and Te,idiv near the separatrix, and to produce the partial detachment. Integrated designs of the water cooled divertor target, cassette and coolant pipe routing have been developed in both EU and JA, based on the tungsten (W) monoblock concept with Cu-alloy pipe. For year-long operation, DEMO-specific risks such as radiation embrittlement of Cu-interlayers and Cu-alloy cooling pipe were recognized, and both foresee higher water temperature (130-200 °C) compared to that for ITER. At the same time, several improved technologies of high heat flux components have been developed in EU, and different heat sink design, i.e. Cu-alloy cooling pipes for targets and RAFM steel ones for the baffle, dome and cassette, was proposed in JA. The two approaches provide important case-studies of the DEMO divertor, and will significantly contribute to both DEMO designs

Progress on reliability of remote maintenance concept for JA DEMO

JA DEMO selected the vertical maintenance scheme as the primary maintenance option. In order to improve reliability of Remote Handling (RH) and plant availability, update of in-vessel transferring mechanism of the segment, pipe unit structure on maintenance port were carried out. The RH equipment is composed of end-effectors (grippers) for the banana segment, a power manipulator, a telescopic guide and a carrier. The pipe structure in vertical maintenance port was updated considering the thermal expansion, easy handling (short maintenance time) and alignment of welding groove. Based on estimation of maintenance time, four works in parallel will be required for plant availability reachable for commercialization (>60%)

Development of poloidal horseshoe limiter concept for JA DEMO

In the development of a DEMO fusion reactor, mitigating the heat load deposited on the breeding blanket at steady-state operation is one of the challenging issues because of the localized high heat flux brought by charged particles and low thermal conductivity of the Reduced Activation Ferritic/Martensite (RAFM) steel structural material. This study proposes a concept of limiter for JA DEMO; a horseshoe-shaped limiter which is continuous poloidally in the first wall except for the divertor area, and discretized toroidally. The limiter is protruded from the first wall to shade the blankets from the charged particles, which move along the magnetic field lines. To enhance the heat removal capability and neutron irradiation tolerance, a tungsten mono-block using RAFM heat sink was applied as the plasma-facing unit (PFU) on the limiters. The heat removal capability of the limiter’s PFU was evaluated as 4.6 MW/m2. The limiter shape satisfying this value was determined basing on the utilization of heat load evaluation with three-dimensional magnetic field line tracing. Three limiter shape parameters; protrusion height, protrusion width and the first wall set-back, were investigated to understand the variation of charged particle load distribution. Both cylindrical and box-shaped blanket concepts were considered, and design windows for the limiter shape were acquired to discuss the optimum parameter. The optimum limiter shape fulfilling the requirement of steady-state operation has been defined, and the limiter’s occupied area amounts to 1.2 % of the FW area for both blanket options

Progress in design and engineering issues on JA DEMO

原型炉設計合同特別チームのこれまでの成果である日本の原型炉概念(JA DEMO)の基本設計として、核融合科学技術委員会が示した原型炉の目標（① 数十万kWを超える定常かつ安定した電力、② 実用に供しうる稼働率,③ 燃料の自己充足性を満足する総合的なトリチウム増殖）を満足する定常炉概念と今後のR&D課題について報告した。特に主な成果として、発電プラント設備を含む原型炉全体像の構築とそれ基づく正味電気出力の明確化、耐圧性とトリチウム生産性を両立するトリチウム増殖ブランケットの概念設計、及び原型炉の安全性確保に向けた分析した想定起因事象とその緩和方策について報告した。第28回 国際原子力機関（IAEA）核融合エネルギー会議（IAEA-FEC2020

Progress of divertor design concept for Japanese DEMO

Power exhaust scenario for the feasible DEMO plasmas and the divertor design have been studied with a high priority in the steady-state Japanese (JA) DEMO with the fusion power of 1.5 GW-level and the major radius of 8 m-class. The power exhaust concept requires large power handling in the SOL and divertor, i.e. Psep~250 MW, and Psep/Rp~30 MWm-1 corresponds to 1.8 times larger than ITER. Long leg divertor (Ldiv = 1.6 m; 1.6 times longer than ITER) was proposed as a reference design. SONIC simulation demonstrated that the peak heat load on the target (qtarget) was reduced to less than 10 MWm-2 under the partially detachment with large radiation fraction of (Pradsol+Praddiv)/Psep ~0.8. A design concept of the monoblock target and cooling water pipes for the JA DEMO was proposed in 2016 [1]: two different water-cooling units, i.e. 200C, 4MPa pressurized water in CuCrZr pipe and 290C, 15MPa pressurized water in Reduced Activation Ferritic–Martensitic steel (F82H) pipe, are used. The heat exhaust unit with the CuCrZr pipe can be applied near the strike-point (0.8 m) for the high heat load region, while the replacement is expected every 1-2 years due to the maximal irradiation dose on the CuCrZr pipe of 2 displacement-per-atom (DPA). Recently, cassette structure for the DEMO divertor was designed to incorporate the heat exhaust units and coolant pipes. One cassette covers the toroidal area of 7.5, and 3 cassettes are replaced from a lower maintenance port (total 16 ports). The cassette design is consistent with reduction in the fast neutron flux to protect the vacuum vessel and replacement of the inner and outer heat exhaust units of the CuCrZr pipe. The cassette structure consists of F82H, and the total thickness of 25 cm can reduce the fast neutron flux efficiently by arranging two lines of puddles for the pressurized water with the path length ratio of 7:3 for F82H and water, respectively. The water flow (1m/s) in the puddles removes the total nuclear heat of 0.7 MW in one cassette (totally 32 MW for 48 cassettes). Heat transport analyses of the W-monoblock and CuCrZr-pipe was performed in the three-dimensional (3-D) modeling by using the ABAQUS finite element method (FEM) code, considering the monoblock geometry (shaped target surface is used to protect the leading edge) and the heat flux components (radiation power, neutral flux and nuclear heat as well as plasma heat load along the field line) given by SONIC and MCNP-R simulations. Maximum temperature on the W-surface appears near the downstream edge in the plasma-wetted area. The critical operation temperature of 1200C, i.e. W-recrystallization, corresponds to the total peak heat load of 13.5 MWm-2, which is 1.8 times higher than the result simulated by SONIC (7.5 MWm-2). The maximum temperature of the CuCrZr pipe is 351C, and mechanical toughness of the cooling pipe is also near critical against thermal fatigue. Elasto-plastic analysis of the displacement and thermal stress on the W-monoblock and CuCrZr pipe under the higher heat load have been performed, and the results are presented.Third Technical Meeting on Divertor Concept

Progress in design and engineering issues on JA DEMO

原型炉設計合同特別チームのこれまでの成果である日本の原型炉概念(JA DEMO)の基本設計として、核融合科学技術委員会が示した原型炉の目標（① 数十万kWを超える定常かつ安定した電力、② 実用に供しうる稼働率,③ 燃料の自己充足性を満足する総合的なトリチウム増殖）を満足する定常炉概念と今後のR&D課題について報告した。特に主な成果として、発電プラント設備を含む原型炉全体像の構築とそれ基づく正味電気出力の明確化、耐圧性とトリチウム生産性を両立するトリチウム増殖ブランケットの概念設計、及び原型炉の安全性確保に向けた分析した想定起因事象とその緩和方策について報告した。第28回 国際原子力機関（IAEA）核融合エネルギー会議（IAEA-FEC2020