Status of Technical Activities of ITER in Korea H.G. Lee, S. Cho, H.J. Ahn, K.J. Jung, and ITER Korea Team ITER Korea (KODA) National Fusion Research Institute KSTAR Conference 2014 24-26 February 2014, Mayhills Resort, Gangwon, Korea
< Contents > < Contents > I ITER Project and its Challenges II ITER Activities in KODA III Fusion Development in Korea IV Summary 2
ITER Project and its Challenges I 3
Fusion Research and ITER Project Fusion Research and ITER Project 20 th Century: US-EU-RF-JA lead fusion studies; ITER Project: International research project As a result, the fusion research reached on the final in participation of world demonstration to assess the scientific and technological leading scientists & engineers feasibility of fusion energy realization. K. Lackner Fusion Spherical tokamaks Goal Strongly shaped REACTOR Divertor High field Superconducting DEMO Compression DT operation Start operation 4
ITER Project ITER Project ITER is a necessary step on the way to commercial fusion reactor ; ITER will demonstrate the feasibility and integration of science and technologies , and safety features for a fusion reactor; The self-sustained D-T burning plasma in ITER will generate 500 MW which is 10 times more power than it receives; ITER enterprise will create a new collaborative culture and standard solving energy and environmental problems and contributing to the world peace; All of the intellectual properties obtained belongs equally to all seven Members. ITER Site under Construction Tokamak Pit R=6.2 m, a=2.0 m, I p =15 MA, 1.4 km x 1 km B T =5.3 T, M=23,000 tons 5
ITER Members (CN, EU, IN, JA, KO, RF, US) ITER Members (CN, EU, IN, JA, KO, RF, US) 6
Challenges of ITER Project Challenges of ITER Project International Enterprise • A number of stake-holders: there are too many different views on ITER. • In-kind procurement system: there are too many interfaces. • Roles and responsibilities of IO and 7 DAs are not clearly defined. • Quality: how to control and manage the quality of in-kind components/system. First-of-a-kind Fusion Device -> Technical Challenges • Design of key components is not completely frozen until now. It is still on-going for 6 years after the start of construction. • There are no explicit lessons learned on a number of technical issues. The mechanism of decision making is too late. Nuclear Fusion Reactor -> Safety Requirements • Nuclear safety issues: IO should respect the French Nuclear Regulations. • Safety requirements come later since site-selection after FDR 2001 baseline. 7
ITER Design and its Review ITER Design and its Review History of Design • 1988 - 1991 (CDA) • 1992 - 1998 (EDA) • 1999 - 2001 (EDA-Extended) FDR 2001 Baseline • 2001 - 2006 (CTA and ITA) Cadarache as ITER site (28 June 2005) JIA (Nov. 2006 / DG in office) In-kind sharing was undertaken on the basis of the FDR2001 baseline/cost. • Nov. 2006 - Nov. 2007 (ITER Design Review) Significant Design Changes (scope/schedule/cost) • 24 October 2007 (Entry-into-force of JIA) Nuclear Safety Requirements <- Fukushima earthquake (April 2011) • July 2010 (new Baseline 2010) 8
Major Issues during ITER Design Review (2007) Major Issues during ITER Design Review (2007) • Physics : ELM suppression (H-mode) and vertical stability • Safety : No Carbon in the Tritium Phase (W) + Tritium Management • Buildings : Re-optimization of layout of tokamak-complex and hot-cell buildings • Magnets : No major changes. PF coils may be changed for plasma control (minor modifications of PF 2&6). Coil cold test is the biggest issue, and a cost driver. • Vacuum Vessel : No major changes (C&S: RCC-MR 2007) • H&C Drive : NBTF (Padua, Italy) and RF antenna (2 x 10 MW) • Tritium Plant : Complete re-design of layout, but no major cost changes • In-Vessel Components : Several adaptations and changes: - Blanket attachment and water cooling manifold - Use of Tungsten in Tritium phase for divertor -> Initial usage of W for divertor 9
High Priority Technical Issues (STAC-2, Nov 2007) High Priority Technical Issues (STAC-2, Nov 2007) Technical Issues Contents Cost (MEuro) 1 Vertical Stability -Stabilize plasma vertical position (<1s) Shape Control / -Increase capability of plasma shape In-vessel Coil: 2 Poloidal Field Coils control and flux swing in Ohmic 220.22 operation by PF (2,5,6) coils (45 kA → Flux Swing in Ohmic PF Coil: 31.23 3 55 kA) and PF 2&6 coils (minor Operation and CS CS Coil: 30.40 modifications) 4 ELM Control -Control & mitigate damage due to ELMs 5 Remote Handling IVT: 28.71 -Study and clarify RH issues 6 Blanket Manifold RH 7 Divertor Armour Strategy -When replacement of W (DT) with CFC? 8 Capacity of 17 MA Discharge -17 MA discharge → Why? 9 Cold Coil Test -Risk mitigation, Nb3Sn (TF), NbTi (PF) 89.85 10 Load Condition on Vacuum -According to JET exp., EM load is Vessel / Blanket higher than the previous specification. 11 TBM Strategy 12 Hot Cell Design -Need ~ 2 times space 111.80 13 H&CD Strategy, Diagnostics -Test facility for NBI NBTF: 92.86 and Research Plan -Test facility for Port Plug Plug TF: 21.00 10
VS Coils (15 MA, Q=10) (0.6 < li < 1.2) VS Coils (15 MA, Q=10) (0.6 < li < 1.2) Stabilization of plasma vertical position H-mode ITER-like discharge li~0.75-0.85 (15 MA, Q=10) (0.6 < li < 1.2) (VS = ± 6 kV) Key issue: o hmic start-up high li end of burn high li (~1.0-1.2) - Reference design (0.7< li <1.0) is to limit the ability to optimize the reference scenario. Passive stabilization improvements (by linking blanket modules for ring circuit) can improve the VS performance to meet the ITER control requirements It was refused due to magnetic measurement (shielding effect). Active stabilization improvements (VS1 = ± 9 kV, VS2 = ± 6 kV for CS2U & 2L) can meet control requirements only over restricted range in l i . it was refused due to safety operation of PS. Exploitation of RMP coils for ELM control can provide required control capability over range in l i expected in current ramp-up and flat-top ( by producing the fast radial field ). Configure in-vessel RMP coils and include necessary additional power supplies. Install in-vessel VS coils for vertical stabilization → (PS: 0.9 kV, 240 kAt) 11
ELM Control is Essential in ITER (IAEA-FEC-23) ELM Control is Essential in ITER (IAEA-FEC-23) uncontrolled determined by ELM physics • DW ELM controlled 0.5 MJm -2 • Material damage avoidance + ELM physics required DW ELM controlled ~ 0.7 MJ (15 MA, Q DT = 10, A ELM = A s.s. ) DW ELM ITER q 95 = 3 20 Ucontrolled ELMs Controlled ELMs A ELM = A s.s. Controlled ELMs A ELM = 4 A s.,s ELM (MJ) 15 (IAEA-FEC-23, 10 A. Loarte) W 5 0 4 6 8 10 12 14 16 I p (MA) Uncontrolled ELMs - Operation is limited to Ip ≤ 6 - 9 MA. • Uncontrolled ELM operation with low erosion up to I p = 6.0–9.0 MA depending on A ELM (DW ELM ) No ELM damage for initial H-mode operation in ITER. 12
ELMs Control by RMP Coils ELMs Control by RMP Coils Key issue: Type I ELMs H-mode in ITER (risk: > 0.3% > 1 MJ) Type I ELMs in ITER (15 MA, Q=10) will exceed the acceptable power density (thermal load) on the divertor target by a factor ~20. (~20 MJ) - This will reduce target lifetime, pollute the plasma with impurities and cause disruptions. ELM mitigation/suppression is so critical for ITER. - Ongoing experimental and theoretical work is focused on both pellet pacing control and RMP (ELMs) coils control. - Pellet pacing experiment on AUG is succeeded in increasing ELM frequency by ~2. Need to increase the frequency by 10 to 20. In ITER, ~100 pellets in ~3 sec. In the end, in-vessel ELM coils (27) will be installed . R&D & engineering design by PPPL (US) and proto-type fabrication by ASIPP (CN) are being carried out. 13
Tungsten Divertor Armour in ITER Tungsten Divertor Armour in ITER The IO concluded that sufficient progress has been made in design and technological development for implementation of the full-tungsten divertor. Optimize tilting of Vertical Targets and Dome to protect inter-cassette leading edges Outer baffle shaping to mitigate W melting at downward VDE impact Individual monoblock shaping in high heat flux areas to protect all leading edges 14
ITER Research Plan: H-mode Power Threshold ITER Research Plan: H-mode Power Threshold • The latest H-mode threshold power scaling for deuterium plasmas: 0.72 B 0.8 S 0.94 thresh 0.05 n (Y Martin, HMW-2008) P e T • The isotope dependence based on JET results in H, D, and DT indica tes that P th 1/A for hydrogen isotopes Possible helium H-mode access half-field/ half current H-mode development DT H-mode access Q=10 Full-field/ full current H-mode development No H-mode access in H at full field No H-mode access in D for full Q=10 simulation • Note: margins may be required for (i) core radiation and (ii) access to good confinement (H 98 = 1) * Note: JET (P LH,He ~ 0.7 P LH,H ) 15
ELMy H-modes of He Discharge ELMy H-modes of He Discharge He Type I ELMy H-modes are key to development of ITER Research Plan. AUG, Ryter et al DIII-D, Gohil et al • Recent experiments on H-mode access in helium indicate that P th,He ~ 1 – 1.5 x P th,D - offers better option for H-mode access in non-active phase - but ELM behaviour and control still require R&D 16
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