Future Fusion Reactor Design Strategy and Experimental Plans

1. DCLL TBM: Comments on Design Strategy, andExperimental plans, schedule and required diagnostics Compiled by Neil Morley for the TBM Meeting May 10, 2006

2. My suggestions for near term Design focus (slight modification of clement’s) • Basic TBM fabrication and mechanical design • FS Fabrication, requirements and limitations • Application of FW heat transfer enhancement techniques • Routing and uncertainties in He and PbLi flow distribution (supply, manifolds, routing) • Basic diagnostics placement and feedthroughs • Demonstration and optimization of DCLL DEMO characteristics • Desired DEMO operating point(s), loads, parameter ranges • Desired DEMO material properties targets • Scaling and geometric similarity of TBM experiments from DEMO conditions to ITER D-T • Initial testing in ITER H-H • ITER Safety and Reliability Requirements • Licensing envelope and accident analysis • Allowable volume of PbLi • Design rules • Disruption response analysis

3. Proceed with most critical R&D • Thermofluid / MHD modeling tools progress and preparation of manifold MHD experiments (Smolentsev) • 1st Generation FCI development and characterization (Katoh) • Material System Compatibility literature and existing data review (no presentation) • FS fabrication development using HIP, welding development, investment casting assessment study, test methods and ITER database (Rowcliffe/Kurtz) • Virtual TBM – foundations and technical planning (no presentation) • Diagnostics planning and survey (I will begin to discuss)

4. Current ITER Operational Schedule

6. ITER’s Description of Early ITER Phases Integrated Commissioning - This phase completes the construction of ITER by ensuring all systems operate together and includes the preparation of the machine to attain the first hydrogen plasma. Hydrogen Phase - This phase allows full commissioning of the tokamak system in a non-nuclear environment without depending on fully-remote handling. • Major aspects of the full DT discharge scenario can be checked, including: • plasma current initiation, current ramp-up & ramp-down, formation of a divertor configuration, • control of and loads due to disruptions or vertical displacement events, • access to H-mode and adequacy of heating installed, • operational limits on density, beta, safety factor, • requirements/capability of steady state operation. • The peak heat flux onto the divertor target will be of the same order of magnitude as for the full DT phase. • Some important issues cannot be fully tested in this phase. These include: • evaporation of the divertor target surface expected during a disruption, • effects of neutron irradiation of the in-vessel materials, • alpha-particle heating of the plasma. • Although there are no neutrons in this phase, test blanket module electromagnetic and hydraulic tests can take place and give very useful information further iterated system designs to be installed for DT operation. • The actual length of this phase depends on the merit of the ongoing experimentation with regard to the later DT operation, in particular the ability to achieve good H-mode confinement with a suitably high plasma density.

7. Loading conditions in H-H and D-T Phases

8. Answers needed on Plasma Operation during H-H phase to plan testing • Will plasma pulses be grouped into operating campaigns, with ITER not operating for extended periods? • Will disruptions and VDEs be intentionally triggered at known times with known, increasing intensity? • Will “actual length” likely be shorter? Or longer?

9. DCLL H-H Phase Testing Strategy & Overall Goals DCLL TBM “Prototype” testing Strategy : • Gather a database of operational information via repeated parameter (flowrates, temperatures, plasma conditions) scans • Operate TBM empty of PbLi until early disruption tests demonstrate TBM and system integrity. • Establish testing capability, performance baseline and operational experience of the TBM and ancillary systems • Integration of control systems and diagnostics with ITER systems • Demonstration of required port integration and remote handling procedures • Measurement of thermal time constants and heat loss, control time constants • Testing heating/filling/draining/remelting and accident response procedures • Validate general TBM structure and design • Mechanical response of the TBM structure to transient EM loads • Determine ferromagnetic and MHD flow perturbation of ITER fields • Measure thermal and particle load effects on plasma facing surface (Be) and FW structure/heat sink • Perform initial studies of MHD effects and Flow Channel Insert performance • MHD flow distribution and pressure drop in toroidal field and toroidal + poloidal field • FCI performance changes as a function LM exposure time and loading from EM events • Map ITER field in TBM area

10. Strawman Testing Schedule in Early H-H

11. Strawman Testing Schedule in Late and Post H-H

12. Proposed transducers needed to meet H-H phase TBM testing goals Check existing plasma exp and ITER basic machine

13. Impact on Proposed Testing on Design of H-H Phase DCLL TBM and systems • H-H TBM • should include FCIs and be • based on the fabrication technology and materials proposed for the D-T phase – both structure and FCIs • Number and position of diagnostics must be determined concomitant with the analysis and design • Feed-through for diagnostic lines in the TBM • Attachments and lead routing inside the TBM • Helium coolant systems and PbLi systems • Variable flowrate (bypasses of VFD) and inlet temperature control (Heaters and/or HX) • Leak checking system (tracer gas?) • Include diagnostics and control elements

14. ITER’s Description of ITER D-D phase plasma operation • In this phase, neutrons will be produced, and tritium will be produced from DD reactions. • Part of this tritium will then be burnt in DT reactions. • the activation level inside the vacuum vessel will not allow human access after a few deuterium discharges with powerful heating. • the capacity of the heat transfer system (except for the divertor and heating devices) could initially be minimal, and demand for the tritium processing system would be very small. • later in the phase integrated DT commissioning can take place, with short pulses at high fusion power. • The major achievements would be as follows: • replacement of H by clean D plasma; • confirmation of L-H threshold power and confinement scalings; • establishment of a reference plasma (current, heating power, density, detached/semi-detached divertor, ELMy H-mode, etc.); • particle control (fuel/ash/impurity/fuelling/pumping); • steady-state operation with full heating power; • finalisation of nuclear commissioning with a limited amount of tritium; • demonstration of high fusion power comparable to the nominal value for the full DT burn, for a short time. • Some information can be provided in this phase on test blanket neutronics behaviour, allowing optimisation of the designs for later DT operation.

15. Nuclear Field / Tritium Production (N/T) TBM • Purpose: • database of neutron field measurements for various types of ITER discharges and conditions • characterize tritium production rate (TPR), and nuclear heating rates. • FW He cooling and tritium implantation • Testing of tritium processing and control equipment – 1st permeation data • Design: • Similar design and structure as the Integrated-TBM • Rabbit-style tube system for deploying/retrieving activation foils into several location in the TBM • Deploy tritium processing system • Tritium diagnostics (Li glass, other) • Nuclear heat (micro-calorimeter, other) diagnostics • Testing during DD and early DT phase: • ~2 years in-ITER • Required operational conditions still to be determined

16. ITER’s Description of ITER Low Duty D-T phase plasma operation • Fusion power and burn pulse length will be gradually increased until the inductive operational goal is reached. • Non-inductive, steady-state operation will also be developed. • Test blanket modules will begin to accumulate results in a situation resembling their operating environment, allowing fine tuning of the designs, and a reference mode of operation for that testing will be established.

17. Thermofluid / MHD (T/M) TBM • Purpose: • thermal and electrical insulation properties of the FCI with nuclear heating • FCI failure rates and effects with nuclear heating • tritium permeation through FCIs and into helium coolants • Pbli natural convection effects on thermal performance • Initial data on activation products and chemistry control • Design: • Aspects of TBM itself still TBD based on ongoing R&D and scaling • 1 module or multiple submodules on a strongback? • Multiple channel sizes and FCIs • Temperature, electric potential diagnostics • Calorimetry • Tritium counting diagnostics • Testing in low duty DT phase • ~2 years in-ITER • moderate temperature (<400-500C) operation of the TBM with PbLi, • possibly with short excursions with PbLi above 500C

18. Description of ITER High Duty D-T phase plasma operation • Improve overall performance, emphasise • testing of components and materials with higher neutron fluences, • aim for high availability • further improved modes of plasma operation. • The implementation and length of this phase will be depend on the results from the preceding three phases and assessment of the merits and priorities of programmatic proposals. • Whether and when to incorporate tritium breeding during this phase will be decided on the basis of the availability of tritium from external sources, the results of breeder blanket testing, and experience with plasma and machine performance. • such a decision would lead to a non-operating period of about 2 years while the blanket system is installed in the outboard plasma region, as provided for in the design and initial installation • the opportunity would undoubtedly also be taken to upgrade ancillary equipment at that time.

19. Integrated Operation (I) TBM • Purpose: • Explore longer term Integrated operation of the system while thermally simulating as near as possible DEMO operation conditions including small accumulation of radiation damage in FCIs and RAFS joints • test improved FCI materials • corrosion in thermal/chemical environment • tritium permeation • Investigate/improve online tritium recovery from PbLi and He streams • Investigate/improve online PbLi purification systems • Design • Preliminary design outlined in DDD (shown by Wong and Dagher) • Testing in high duty DT phase • ~3 years in-ITER • Continuous operation in long campaigns looking for changes in performance and failures