Overview of the ARIES Fusion Power Plant Studies Farrokh Najmabadi IAEA Technical Committee Meeting on Fusion Power Plant Studies March 24-28, 1998 Culham, United Kingdom

The ARIES Team Has Examined Several Tokamak and non-Tokamak Power Plants in the Past 10 Years TITAN reversed-field pinch (1988) ARIES-I first-stability tokamak (1990) ARIES-III D- 3 He-fueled tokamak (1991) ARIES-II and -IV second-stability tokamaks (1992) Pulsar pulsed-plasma tokamak (1993) SPPS stellarator (1994) Starlite study (1995) ARIES-RS reversed-shear tokamak (1996) ARIES-ST spherical tokamak (in progress)

ARIES-RS

No public evacuation plan is required: total dose < 1 rem at site boundary; Generated waste can be returned to environment or recycled in less than a few hundred years (not geological time-scale); No disturbance of public’s day-to-day activities; No exposure of workers to a higher risk than other power plants; Closed tritium fuel cycle on site; Ability to operate at partial load conditions (50% of full power); Ability to maintain power core; Ability to operate reliably with less than 0.1 major unscheduled shut-down per year. Top-Level Requirements for Commercial Fusion Power Plants Extra Above requirements must be achieved consistent with a competitive life-cycle cost of electricity goal.

Tokamak Research Has Been Influenced by the Advanced Design Program  p /A (  Bootstrap current fraction)  A/S (  Plasma  “Conventional” high-  tokamaks (Pulsed operation) 2nd Stability high-  tokamaks (Too much bootstrap) Advanced tokamak (Balanced bootstrap) PU: Pulsed Operation SS: 2nd Stability FS: 1st Stability, steady-state RS: Reversed-shear Current focus of tokamak research

Design ClassFeaturesDistinct Issues Ferritic SteelLimited efficiency Ceramic or LM breederLarge data baseLimited heat flux He, LM, or H 2 O coolantLarge followingFerromagnetism Vanadium-alloyDatabase/industry Li or ceramic breederHigh performanceHigh unit cost He or Li coolantLow afterheatCompatibility Waste disposalCoatings SiC compositesMaterial form& Ceramic breederVery high performance properties He coolantExcellent safety &High fabrication Waste cost Engineering Design Options were Assessed Based on High-performance Structural Material

ARIES-RS Design Emphasized those Features that Maximize Attractiveness  Quantitative goals defined in the form of top-level requirements –Evolved from frequent interaction with customer.  Reversed shear mode of plasma operation: –High , high    high I BS, transport suppression.  High-performance self-cooled lithium with vanadium in high- temperature zones: –Based on ARIES-II with numerous cost-saving measures.  Availability a major engineering trust: –Design for full-sector maintenance with detailed analysis.

Major Parameters of ARIES-RS Design Aspect ratio4.0 Major toroidal radius (m)5.5 Plasma minor radius (m)1.4 Plasma elongation1.7 Plasma triangularity0.5 Toroidal  Electron density (10 20 m -3 )2.1 ITER-89P scaling multiplier2.3 Plasma current11

Major Parameters of ARIES-RS Design Current-drive power to plasma (MW)81 On-axis toroidal field (T)8 Peak field at TF coil (T)16 TF-coil ohmic losses (MW)-- Peak/Avg. neutron wall load (MW/m 2 )5.4 / 4 Fusion power (MW)2,170 Gross electric power (MW)1,200 Recirculating power fraction0.17 Cost of electricity (mill/kWeh76

ARIES-RS is a conceptual 1000MWe power plant based on a Reversed-Shear tokamak plasma

Key Performance Parameters of ARIES-RS

Critical Physics Issues for Advanced Tokamaks  Wall-stabilization of kink modes.  Current drive near plasma edge and especially at mid- plasma.  Current drive power is sensitive to constraint imposed by the divertor:  High separatrix density;  Impurity injection to radiate the plasma energy;  Achieving the necessary density and temperature profiles.  Start-up, access, and

ARIES-ST

The ARIES-ST Study - Goals & Schedule  The ARIES-ST study is a two-year project to investigate the potential of spherical tokamaks as commercial power plants as well as vehicles for fusion development.  The ARIES-ST study began in Jan The effort has been focused on ST power plants. We have emphasized understanding the trade-off and identifying issues that have to be resolved.  The ARIES-ST study will be completed in Fall The research reported in this meeting represents a progress report as in many areas design work is not completed.

The ARIES-ST Study - Background  Theoretical and experimental studies indicate that the MHD performance of a tokamak plasma is substantially improved with decreasing aspect ratio.  Tokamak power plants with superconducting TF coils, however, tend to optimize at A  4 as the gain in  at lower A is offset by gains at higher A through:  Higher on-axis field for a fixed maximum field at the coil;  Lower current-drive power (because of lower plasma current);  Engineering advantages of additional available space.  Question: What is the optimum regime of operation for tokamaks with resistive coils:  for power plants (Joule losses in TF is critical);  for fusion development or non-electric applications (Joule losses in TF may not be as critical).

Key Physics Issues for Spherical Tokamaks  Because of low aspect ratio, the area in the inboard is limited. A resistive TF coil is probably the only option because of the lack of space for a shield for a cryogenic superconducting coil.  In order to minimize the Joule losses in the TF coils (mainly the center-post), MHD equilibria with very high  are required.  Because there is no room for a central solenoid, steady- state operation is mandatory. Because of large plasma current, only MHD equilibria with almost perfect bootstrap alignment would lead to a reasonable current- drive power.

Key Physics Issues for Spherical Tokamaks  Because of unique magnetic topology, on-axis current drive with RF techniques is difficult. Current drive for profile control as well as start-up are additional challenges.  The divertor problem is more difficult than conventional and advanced tokamaks (higher P/R).  Extrapolation of present confinement data base (scaling) to fusion regime is questionable.

Key Engineering Issues for Spherical Tokamaks  The small area available for the inboard legs of the TF coils (center-post) make the design of center-post challenging.  Potential advantages of spherical tokamaks (compact and high wall load) make the engineering of fusion core difficult:  Because of large recirculating power, a highly efficient blanket design is essential;  Water-cooled copper coils further narrow the options;  High heat flux on in-vessel components further narrows the options;  Highly shaped components (tall and thin) make mechanical design difficult.  Maintenance of the power core should include provisions for rapid replacement of center-post.

Spherical Tokamaks Are Quite Sensitive to Physics/Engineering Trade-off  The physics and engineering trade-off are most evident in determining the inboard radial built:  Smaller radial built  improved plasma performance;  Larger radial built  engineering credibility;  Every centimeter counts!  Challenge: maximize physics performance while maintaining a credible design.

Major Parameters of ARIES-ST Strawman Aspect ratio1.6 Major toroidal radius (m)3.3 Plasma minor radius (m)2.1 Plasma elongation3.2 Plasma triangularity0.57 Toroidal  Electron density (10 20 m -3 )3.0 ITER-89P scaling multiplier2.7 Plasma current32

Major Parameters of ARIES-ST Strawman Current-drive power to plasma (MW)57 On-axis toroidal field (T)2.8 Peak field at TF coil (T)11.5 TF-coil ohmic losses (MW)871 Peak/Avg. neutron wall load (MW/m 2 )8.2 / 5.4 Fusion power (MW)4245 Gross electric power (MW)2204 Recirculating power fraction0.55 Cost of electricity (mill/kWeh)111

TF Coil System Is Designed for Vertical Assembly Water-cooled center-post is made of DS GlidCop AL15. Outboard TF coil form a shell to minimize mechanical forces. Center-post is connected to the TF shell through a tapered joint on the top and sliding joints at the bottom. Insulating joint is located at the outboard mid-plane where the forces are smallest. Another TF joint is provided for vertical maintenance of the power core.

Vertical Maintenance from the Bottom Is Preferred Reduced building height & size. Radioactive material are confined to the maintenance area. More accurate positioning with lifts compared to cranes.

The Fusion Core Is Replaced as a Unit

MHD Equilibrium and Stability  The MHD stability of ST discharges for a wide range of aspect ratio, elongation, triangularity, and kink wall location was examined (with ~ 99% bootstrap fraction).  There is a high leverage to operate at high elongations (and high  ) in order to achieve a high . It appears that operation at  ~ 3 is possible. Detailed work in quantifying the feed-back power necessary for vertical stabilization of high-elongation ST plasmas is on-going.  Low-A free-boundary equilibria is unique and difficult to calculate:  Strong B variation;  Strong plasma shaping (  ~ 3,  ;  High  p (~2) and low l i (~0.15).

Free-Boundary Equilibria with Different Elongations

Current Drive  High-frequency fast wave (HFFW) can drive the current in the mid-plasma efficiently.  It appears that LFFW is the only plausible RF technique that drives current near the axis on high-  ST plasmas:  Because  pe /  ce >>1, EC and LH waves cannot access the plasma center.  HFFW does not penetrate to the center because of strong electron and/or ion damping;  ICRF fast wave suffer strong electron and  /ion damping.  LFFW requires a large antenna structure for a well-defined spectrum (   ~ 14 m). It generally has a fairly low current-drive efficiency.  We are also exploring NBI as an alternative options.

Divertor  The divertor problem is more difficult than conventional and advanced tokamaks (higher P TR./R ).  To reduce the heat flux to a manageable level, a large fraction of the plasma power has to be radiated:  Radiative mantle;  Impurity radiation in the divertor channel.  Impact of finite edge density and impurities on the MHD/current drive is under investigation.  This approach mainly transfers the divertor problem to the first wall!

Center-post Design -- Inboard Shield  A 20- to 30-cm thick inboard shield is required:  To allow center-post to meet low-level waste disposal requirement;  To reduce nuclear damage to the conductor;  To limit Joule losses due to neutron-induced transmutation;  To reduce nuclear heating in the center-post conductor;  To improve power balance by recovering high grade heat from shield;  To prolong the center-post life time to up to 3 years (same as first wall) in order to minimize impact on availability and replacement cost.

Transmutation of Cu Changes the Center-post Resistivity  Dominant Cu transmutation products are Ni, Zn, and Co  64 Ni and 62 Ni dominate the change in resistivity Resistivity changes with a 30-cm, 80% dense Ferritic Steel/He shield

Electrical Design of the Center-post  Leading conductor material is Glidcop AL-15.  It has adequate strength, ductility, low swelling, and thermal and electrical conductivities;  Under irradiation, it suffers from severe embrittlement (at room temperature;  Hardening and embrittlement are alleviated by operating above 180C but then it suffers from severe loss of fracture toughness.  Single-turn TF coils are preferred in order to reduce Joule heating  Higher packing fraction;  Reduced shielding requirement (no insulation);  Requires high-current low-voltage supplies with massive busbars.

Mechanical Design of the Center-post  Sliding electrical joints are employed between center-post and other TF legs and bus-bars and TF legs.  They allow relative motion in radial and vertical directions (which minimizes axial loads on the center-post);  They enhance maintainability;  Several design options have been developed and tested successfully.  Center-post is physically separate from other components in order to avoid a complex interface.  We are currently assessing the degree to which the center-post can be flared to reduce Joule losses.

Wedged Center-post Option for ARIES-ST

ARIES-ST Center-post Uses Sliding Joints

Thermal-hydraulic Design of the Center-post  Cry-cooling does not offer major improvement over cooling options at room temperature and above.  Water cooling is the leading option:.  Low-temperature operation (T inlet ~ 35C) minimizes Joule losses but results in sever embrittlement of conductor;  High-temperature (T inlet ~ 150 to 180C) avoids embrittlement but lose of fracture toughness and increased Joule losses are key issue.).  Liquid lithium (both conductor and coolant) is probably the best option for high-temperature operation. However, in addition to many challenging engineering issues, recovery of center-post heating does not offset increased Joule losses.

First Wall and Blanket Options  Design which include solid breeders require a major improvement in the thermal conductivity of solid breeders to handle high wall loads..  Self-cooled Li/V option can handle the high wall load.  The reference blanket design uses ferritic steels as structural material with helium as coolant and LiPb as the liquid breeder. SiC composite fillers are used to achieve a high- coolant outlet temperature and a reasonable power- conversion efficiency.

High-Performance Ferritic Steels Blanket Typically, the coolant outlet temperature is limited to the max. operating temperature of structural material (550 o C for ferritic steels) By using a coolant/breeder (LiPb), cooling the structure by He gas, and SiC insulators, a coolant outlet temperature of 700 o C is achieved for ARIES-ST increasing the thermal conversion efficiency substantially.

Summary -- ARIES-ST  Several key challenging issues confront spherical tokamaks as fusion power plant.  We have proposed some potential solutions.  Some of these constraints are less sever in a non- electricity producing device.  It appears that spherical tokamak power plants do not offer major improvements over advanced high-aspect ratio tokamaks.  In the remainder of this year, we will complete our reference ARIES-ST design and examine potential of spherical tokamaks as vehicles for fusion development.