Summary Tritium Day Workshop

Transcription

1 Summary Tritium Day Workshop Presentations by M. Abdou, A. Loarte, L. Baylor, S. Willms, C. Day, P. Humrickhouse, M. Kovari 4th IAEA DEMO Programme Workshop November 18th, Karlsruhe, Germany

2 Overall Observations Thanks for all the presenters for their efforts in preparing excellent, quantitative, and clear presentations directly relevant to the scope of the tritium day workshop. This enabled us to arrive at much better understanding of the tritium fuel cycle and coupling to plasma physics and fueling schemes, AND quantifying the state-of-theart and the required R&D to: a) Achieve tritium self-sufficiency with high confidence b) Reduce the magnitude of the necessary start-up inventory Quantitative understanding of the dependence on various parameters in the plasma, fueling system, and fuel cycle allows deriving specific quantitative goals for necessary R&D advances

3 Dynamic fuel cycle models were developed to calculate time-dependent tritium flow rates and inventories and required TBR Results show critical importance of plasma fueling and plasma exhaust inventories and processing Simplified Schematic of Fuel Cycle to New Plants Startup Inventory T Storage and Management Fueling System DT Plasma Neutron Blanket Isotope Separation System Fuel Cleanup Vacuum Pumping Divertor/ FW PFC Coolant Water Detritiation System T Waste Treatment Coolant T- Processing T Processing for Blanket depends on design options 3

4 Tritium Start-up Inventory (kg) Technology Advances Tritium inventories depend strongly on tritium burn fraction (f b ), tritium fueling efficiency (η f ), and tritium processing time (t p ) 50 Doubling Time: 5 years 40 t p : tritium processing time Initial Inventory Start-up Inventory t p =24 hrs t p =12 hrs t p =6 hrs t p =1 hr 10 Fusion Power = 3000 MW Reserve time for outage x fraction of tritium plant failing= 0.25 day Inefficiency, ε = 0.01% Blanket mean residence time = 10 days Tritium Burnup Fraction x f (%) Physics x Technology Advances 4

5 Required TBR and achieving T self-sufficiency are strongly dependent on f b x η f, and t p Doubling Time: 5 years Tritium Processing Time 24 hours 12 hours 6 hours 1 hour Max achievable TBR ~ Δ Window for Tritium self sufficiency Tritium Burnup Fraction x f (%) Attaining Tritium Self Sufficiency in DT Fusion Imposes Key Requirements on Physics and Technology. The goal for R & D should be to achieve: T burnup fraction (f b ) x fueling efficiency (η f ) > 5% (not less than 2%) T processing time (in Plasma exhaust/fueling cycle) < 6 hours 5

6 lasma Physics Aspects of Tritium Burn Fraction and Prediction for ITER-1 ITER systems (pellet and gas fuelling) and total throughput (200 Pam -3 s -1 ) provide appropriate flexibility to achieve Q = 10 mission by providing core plasma fuelling, helium exhaust and edge density control for power exhaust (including ELM control) G T burn = 0.35 Pam -3 s -1 + G T fuelling = 100 Pam -3 s -1 G T burn / G T fuelling = 0.35 % Very conservative assumes all fuelling (gas+pellet) done with DT Fueling requirements for edge/power load control and ELM control dominate total throughput and can require up to 130 Pam 3 s -1 requirements for He exhaust are less demanding (~ 40 Pam 3 s -1 out of a maximum of 200 Pam 3 s -1 ) Recycling fluxes and gas puffing expected to be very ineffective in ITER to fuel the core plasma edge and core D/T mixes are decoupled T-burn can be optimized by using only T for core fueling with HFS pellets and D for edge density/power load/elm control G T burn = 0.35 Pam -3 s -1 + G T fuelling = Pam -3 s -1 G T burn / G T fuelling = % A. Loarte 4 th IAEA DEMO Programme Workshop KIT Page 6

7 lasma Physics Aspects of Tritium Burn Fraction and Prediction for ITER-2 Achievable T-burn optimization in ITER depends mostly on two uncertain physics issues : Required edge density (and associated gas fuelling) to achieve power load control (i.e. power e-folding length l p ) Fuelling requirements to achieve ELM control (i.e. throughput associated with pellet pacing for ELM control and pellet+gas fuelling associated with ELM control by 3-D fields) DEMO fuelling and T-burn expected to be similar to ITER except: Pellet deposition more peripheral than in ITER pellet efficiency maybe reduced due to more likely triggering of ELMs after injection of fuelling pellets Higher core radiation and associated edge impurity density can cause pedestal inwards DT pinch which can improve net efficiency of gas fuelling in DEMO compared to ITER A. Loarte 4 th IAEA DEMO Programme Workshop KIT Page 7

8 Fusion Fueling Efficiency Summary A reactor must be designed from the beginning for optimal fueling and pumping efficiency EU DEMO Gas fueling/recycling expected to be highly inefficient, R~0. High fueling efficiency > 50% can possibly be achieved with suitable high speed HFS pellet injection in a tokamak DEMO A stellarator DEMO would also need high speed pellets ELM impact on HFS pellet fueling efficiency remains an open question 8 Calculations of pellet penetration for DEMO conditions show penetration to the pedestal top is possible with HFS injection optimal location under study DEMO 2016 LRB

9 Fueling Efficiency Extrapolation from Deep to Shallow Penetrating Pellets Expected in ITER and DEMO is Highly Uncertain but can Possibly Exceed 50% Dn e (10 20 m -3 ) EU DEMO (Te PED = 6 kev) ITER DEMO DIII-D Ablation without Inward Drift Included 10% 2500 m/s 15% 10% 300 m/s 10% Pedestal 0.2 5% Edge Pellet Depth r Axis Extrapolation from present small tokamaks to ITER and DEMO is highly uncertain, is likely less in DEMO than ITER from more shallow penetration. 0.7 r 9 Ablation profiles (no drift included) show penetration possible to pedestal top with high speed pellets (% density perturbation), but not with slow speed from inner wall. DEMO 2016 LRB

10 Tritium Start-up Inventory (kg) Technology Advances Required start-up tritium inventory based on quantitative understanding from this workshop 50 Doubling Time: 5 years Initial Inventory Start-up Inventory Burn fraction ~ 1.5% HFS fueling efficiency ~ 50% t p ~ 2 6 hrs Start-up inventory ~ Kg Remarkable Progress but Major improvements are still needed!! Loarte & Baylor improvements 0 t p : tritium processing time t p =24 hrs t p =12 hrs t p =6 hrs t p =1 hr Tritium Burnup Fraction x (%) f Physics x Technology Advances 10

11 Confidence in achieving tritium self-sufficiency based on quantitative understanding from this workshop Burn fraction ~ 1.5% HFS fueling efficiency~ 50% t p ~ 2 6 hrs Self-sufficiency changes from unlikely to likely Loarte & Baylor improvements Doubling Time: 5 years Tritium Processing Time 24 hours 12 hours 6 hours 1 hour Tritium Burnup Fraction x f (%) Max achievable TBR ~ 1.15 Window for Tritium self sufficiency Major improvements still needed for attaining Tritium Self Sufficiency with higher confidence level. The goal for R & D should be to achieve: T burnup fraction (fb) x fueling efficiency (ηf) > 5% (not less than 2%) T processing time (in Plasma exhaust/fueling cycle) < 6 hours 11 Δ

12 The Issue of External Tritium Supply is Serious and Has Major Implications on Fusion Development Pathway The start-up tritium inventory required for any reactor or DEMO is a strong function of physics and technology parameters, particularly T burn fraction, fueling efficiency and tritium processing time. - This start-up inventory is ~15-30 kg with current start-of-the-art, and can be reduced to ~8-12 kg if a burn fraction x fueling efficiency of 5% can be achieved. There is no practical external source of tritium available for fusion development beyond ITER (definitely not for multiple DEMOs around the world) - Ontario may be able to supply 5 kg for fusion in 2060, but not 10 kg. - If a reactor is started up around 2060, Canada, Argentina, China, India, Korea and Romania may be able to provide enough tritium for 1 or 2 machines. But, this is highly uncertain. - Start-up with deuterium-rich fuel would delay power production by years and is not economically sensible. - A scheme to generate start-up inventory for DEMO using FNSF has been proposed- merits serious explorations. (may be the only option left?) 12

13 The ITER Tritium Plant design and construction is a major undertaking Scale up by factor of 10 to 20 Comparable to DEMO scale Tokamak Complex Detritiation System Scrubber columns 8 each, 11 m tall 20 m3 Water Detritiation System tank installed (center) Tokamak ring in background IAEA DEMO Workshop, Karlsruhe, Nov 2016 ITER UID: THR7JS Page 13

14 Feasible technologies for each function have been identified for ITER fuel cycle Tokamak Exhaust Processing Permeators, cryogenic molecular sieve, palladium membrane reactor Isotope Separation Systems Cryogenic distillation Storage and Delivery System Uranium hydride beds Detritiation System Catalytic oxidation, scrubber columns Water Detritiation System Electrolyzers, Liquid phase catalytic exchange Analytical System Raman spectroscopy and other analysis techniques IAEA DEMO Workshop, Karlsruhe, Nov 2016 ITER UID: THR7JS Page 14

15 Major themes for tritium technology gap analysis from ITER to DEMO Tritium purification and recycle Safety Proof-of-concept work has been performed, but now this must progress to the next level. ITER will make significant contributions. New technologies will be needed due to impractical scale-ups and to accommodate tritium breeding. Scaling containment/detritiation systems to the next level is proving difficult and expensive. Containment in the extreme DT fusion environment will reveal issues that must be addressed. Tritium breeding and extraction Fundamental experiments are needed. No proof-of-concept experiments have been performed. Full experiments will require tritium source (e.g. from fission neutron irradiation). ITER TBM is planned and complimentary work is needed A large body of work will be required to field a functioning, full breeder blanket on DEMO or pre-demo experiments IAEA DEMO Workshop, Karlsruhe, Nov 2016 ITER UID: THR7JS Page 15

16 C. Day presented a summary of EUROfusion fuel cycle project A number of proposals for innovation in the fuel cycle were proposed. Some of these proposals are in an early stage and would need further evaluation and discussions among experts. The classical DT fuel cycle architecture has been expanded from an all-through to a multi-loop architecture with novel functionalities to address per se the main challenge, i.e. to reduce the integral cycle times and to minimize inventory. Most batch processes at cryogenic temperatures are replaced by continuous processes at non-cryogenic temperatures (cryopumps diffusion pumps; cryogenic viscous compressor liquid ring pump; cryodistillation thermal cycling absorption). The tritium plant systems now feature an outermost loop with classical functionality and an inner semi-continuous isotope separation loop. An additional loop provides separation of the exhaust gas close to the divertor and a shortcut between separated DT and the fuelling systems: Direct Internal Recycling (DIR). The DIR loop is based on novel liquid metal (mercury) technology for vacuum pumping and superpermeation for DT separation. A technical process to implement DIR has been proposed (KALPUREX ) Chr. Day IAEA DEMO WS, Karlsruhe Nov 2016 Page 16

17 C. Day presented a summary of EUROfusion fuel cycle project (cont d) EUROfusion is implementing an R&D programme to advance the novel DEMO fuel cycle towards a conceptual design in the next 10 years. The need for higher fuelling efficiency is addressed in this. This enterprise is following a system engineering approach to make all decisions fully traceable and more easily adaptable, if requirements change. A new concept based on continuous re-injection of exhaust gas in order to increase the recycling coefficient in a metal wall machine and, hence, to increase the burn-up fraction is under investigation. It is essential to implement an integrated and holistic view on the fuel cycle. Examples are the interfaces between inner and outer fuel cycle via the blanket tritium extraction systems and the coolant purification systems, and the divertor which links physics, materials and vacuum technology. Chr. Day IAEA DEMO WS, Karlsruhe Nov 2016 Page 17

18 Tritium Control and Management Tritium control and management will be one of the most difficult issues for fusion energy development, both from the technical challenge and from the public acceptance points of view. The scale-up from ITER to DEMO is orders of magnitude. Why is Tritium Permeation a Problem? Most fusion blankets have high tritium partial pressure. The temperature of the blanket is high ( ºC) Surface area of heat exchanger is large, with thin walls Tritium is in elementary form These are perfect conditions for tritium permeation. The allowable tritium loss rate is very low (~10 Ci/day), requiring a partial pressure of ~10-9 Pa. Challenging! Even a tritium permeation barrier with a permeation reduction factor (PRF) of 100 may be still too far from solving this problem! Barriers have not performed well in irradiated experiments to date 18