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									     FNS related activities
     in the ARIES program


                  M. S. Tillack




Fusion Nuclear Science and Technology Annual Meeting
                   3 August 2010
                       UCLA
                     Topics

1. Design and analysis of plasma-facing
   components: pushing the limits of performance

2. A new systems analysis technique – VASST

3. Evaluation of R&D pathways using
   “Technology Readiness”
   Several W-He divertor designs have been
        studied and evolved in ARIES




                Tapered T-tube

                                             Pin-fin cooling


                                  Finger/plate
                                 combinations




Fingers without W/FS joints
                                                      Tapering reduces temperature and
                                                               increases stress
                                        1400                                                                     450

                                                                                                                                 Non-Tapered 5mm
                                        1375                                                                     425
Maximum Tungsten Tube Temperature (C)




                                                                                                                                 Non-Tapered 1mm

                                        1350                                                                     400
                                                                                                                                 Tapered 5mm
                                                                                                                                 Tapered 1mm




                                                                                              Max Stress (MPa)
                                        1325                                                                     375             Tapered 5mm (λ reduced by 5%)
                                                                                                                                 Tapered 5mm (λ reduced by 10%)
                                        1300                                                                     350


                                                                      Non-Tapered 5mm
                                        1275                                                                     325
                                                                      Non-Tapered 1mm
                                        1250                          Tapered 5mm                                300

                                                                      Tapered 1mm
                                        1225
                                                                                                                 275
                                                                      Tapered 5mm (λ
                                                                      reduced by 5%)
                                        1200
                                               8     9    10    11   12    13       14   15                      250
                                                                                                                       8     9      10    11    12     13    14   15
                                                   Plasma Surface Heat Flux (MW/m^2)                                       Plasma Surface Heat Flux (MW/m^2)

                                               The T-tube divertor is limited by W temperature more than stress,
                                                                      so tapering is helpful
                                    Pin-fin experiments
                                    demonstrated very
                                    high heat removal
5


                     2 mm Bare                  For plate divertor,
                                                 pin-fin array
                     2 mm Pins
    qmax [MW/m2]




                     0.5 mm Bare               • Increases qmax to 18
                     0.5 mm Pins                   MW/m2 at expected
                                                    Re, and to 19 MW/m2
                                                    at higher Re
                                                •   Allows operation at
                                                    lower Re for a given
                                                    qmax  lower
                                                    pressure drop



                            Re (/104)
Fabrication to failure (birth to death) analysis

                  Fabrication Cycle




                                           Operation Cycle,
                                           warm shutdown




       Time-dependent analysis including fabrication, shakedown,
        cycling (warm or cold shutdown), off-normal events.

       Development of reference scenarios for power plants:
        ELM’s, disruptions, VDE’s
       Pushing the limits: beyond 3Sm

  ASME                ½ eue             creep, crack growth,
code (3Sm)          (necking)            irradiation effects
 past                 present                   future


    Self-relieving thermal stress is a large portion of the
     total stress in HHF components.

    Accounting for inelastic behavior (yielding) can
     expand the design window considerably.

    Our goal is steady-state divertor heat fluxes >10 MW/m2
     (>1 MW/m2 for FW’s) and accommodation of transients.
  Elastic-plastic analysis allows for higher
        performance in the first wall


                 Elastic Analysis        Plastic Analysis

               Stress in   Stress in   Stress in   Stress in
Heat Flux      ODS FS       F82H       ODS FS       F82H
                 MPa         MPa         MPa         MPa
0.0 MW/m2
 (390˚C warm    1530         1520        483         370
  shutdown)


1.0 MW/m2       1140         1500        231         202

2.0 MW/m2       1520         1510        315         271

                       stress-free-temperature is 1050 ºC
A modified first wall concept has been examined
                 W pins are brazed into ODS steel plate, which
                  is attached to ferritic steel cooling channels
                 Pins help resist thermal transients and erosion
                 Minor impact on neutronics
                 1 MW/m2 normal, 2 MW/m2 transient applied
                 1 mm FW leads to ratchetting, 2 mm is stable

                                                               0.9

                                                               0.8
                                                                                         F82H



                                      Maximum Plastic Strain
                                                               0.7
           F82H                                                0.6

                                                               0.5
                                                                                        ODS steel
                                                               0.4
                 ODS steel                                     0.3

                                                               0.2
                                                                                             W
                                                               0.1
                  W                                             0
                                                                     0   10   20   30   40   50   60   70   80   90 100
                                                                                    Loading Cycles
External transition joints help alleviate one of
   the most challenging aspects of HHFC’s


                                                 mat’l   ε2d     εallowable
                                                 ODS     0.77%   ~1%
                                                 Ta      0.54%   5-15%
                                                 W       ~0 %    ~1%




2D plane strain analysis
•elastic/plastic, bilinear isotropic hardening
•1050 ºC stress-free (brazing) temperature
•100 cycles 20–700˚C
•3D analysis will be performed next
    Future plans on HHFC (tentative)

   Full elastic/plastic analysis of plate and finger concepts
   3D analysis of joint
   Design integration of pin fins in the plate concept


   Come see us at TOFE:
        “Optimization of ARIES T-tube divertor concept,” J. A. Burke, X. R. Wang,
         M. S. Tillack
        “Elastic-plastic analysis of the transition joint for high performance divertor
         target plate,” D. Navaei, X. R. Wang, M. S. Tillack, S. Malang
        “High performance divertor target concept for a power plant: a combination of
         plate and finger concepts,” X.R. Wang, S. Malang, M. S. Tillack
        “Innovative first wall concept providing additional armor at high heat
         flux regions,” X.R. Wang, S. Malang, M. S. Tillack
        “Ratchetting models for fusion component design,” J. P. Blanchard,
         C. J. Martin, M. S. Tillack, and X. R. Wang
        “Developing a new visualization tool for the ARIES systems code,”
         L. C. Carlson, F. Najmabadi, M. S. Tillack
  Systems analysis in ARIES has evolved
         during the past 2 years
              Fusion Eng. & Design, 85 (2), 243-265, 2010.



     ASC                                         VASST
Determination of an                        Multi-dimensional
optimum design point.                      parameter space scans.

Single-parameter scans                     Large database of physics
around the design point.                   operating points stored.

Difficult to use, maintain                 Graphical user interface.
and modify.
                                           Interactive tool for concept
Non-interactive tool for                   exploration.
self-consistency and costing.
                   Systems analysis flow

     physics               engineering              build out & costing

   Plasmas that             Inboard radial               Top and
   satisfy power              build and               outboard build,
    and particle             engineering                  costing
      balance                   limits


Scan several plasma         Screen physics             Surviving feasible
   parameters to         operating points thru        operating points are
   generate large           physics filters,          built out and costed,
database of physics     engineering feasibility,      graphical display of
  operating points      and engineering filters      parameters (e.g. COE)

                      Filters include e.g.
                         1. Toroidal magnetic
                            fields
                         2. Heat flux to divertor
                         3. Neutron wall load
Number of points
                               VASST GUI v.2
  in database                (Visual ARIES Systems Scanning Tool)
          Blanket database                                           Auto-labeling
               used
                                      Pull-down menus for           Color bar scale
                                      common parameters




                       Constraint parameter can
                          restrict database




          Correlation coefficient

            Save plot as TIFF,
           JPEG, BMP, PNG…


                    Turn on ARIES-AT point
                      design for reference




                               Edit plotting properties
          The code has multiple applications
 scan parameter space, explore tradeoffs (e.g. conservative vs. aggressive),
       describe a design point, and even evaluate research facilities



  We are currently exploring the
4 corners of tokamak design space
  to better understand tradeoffs




                                        Example: optimum size and bN in
                                        the aggressive/aggressive corner
      We adopted “technology readiness levels”
      as the basis for the evaluation of progress
                      Fusion Science & Tech 56 (2) August 2009.

         TRL’s express increasing levels of integration and
       environmental relevance, terms which must be defined
                 for each technology application
TRL         Generic Description (defense acquisitions definitions)               Facilities (mst)
 1    Basic principles observed and formulated.
 2    Technology concepts and/or applications formulated.
 3    Analytical and exptl demo of critical function and/or proof of concept.   Single effects
 4    Component and/or bench-scale validation in a laboratory environment.      Multiple effects
 5    Component and/or breadboard validation in a relevant environment.         Multiple effects
 6    System/subsystem model or prototype demo in relevant environment.         FNSF (PoP)
 7    System prototype demonstration in an operational environment.             Prototype (ITER?)
 8    Actual system completed and qualified through test and demonstration.     Demo reactor
 9    Actual system proven through successful mission operations.               Power plant
                Utility Advisory Committee
     “Criteria for practical fusion power systems”
                        J. Fusion Energy 13 (2/3) 1994.

   Have an economically competitive life-cycle cost of electricity
   Gain public acceptance by having excellent safety and
    environmental characteristics
       No disturbance of public’s day-to-day activities
       No local or global atmospheric impact
       No need for evacuation plan
       No high-level waste
       Ease of licensing
   Operate as a reliable, available, and stable electrical power source
       Have operational reliability and high availability
       Closed, on-site fuel cycle
       High fuel availability
       Capable of partial load operation
       Available in a range of unit sizes
                      These criteria for practical fusion suggest
                       three categories of technology readiness
                         A. Power management for economic fusion energy
                            1.   Plasma power distribution
                            2.   Heat and particle flux management
                            3.   High temperature operation and power conversion
                            4.   Power core fabrication
12 top-level issues




                            5.   Power core lifetime

                         B. Safety and environmental attractiveness
                            6. Tritium control and confinement
                            7. Activation product control and confinement
                            8. Radioactive waste management

                         C. Reliable and stable plant operations
                                                                               cf. GNEP issues:
                            9. Plasma control
                            10. Plant integrated control      •   LWR spent fuel processing
                            11. Fuel cycle control            •   Waste form development
                            12. Maintenance                   •   Fast reactor spent fuel processing
                                                              •   Fuel fabrication
                                                              •   Fuel performance
    Example TRL table: Heat & particle flux handling
                      Issue-Specific Description                                      Program Elements
    System studies to define tradeoffs and requirements on heat flux level,   Design studies, basic research
1   particle flux level, effects on PFC's (temperature, mass transfer).
    PFC concepts including armor and cooling configuration explored.          Code development, applied research
2   Critical parameters characterized.
    Data from coupon-scale heat and particle flux experiments; modeling       Small-scale facilities:
3   of governing heat and mass transfer processes as demonstration of         e.g., e-beam and plasma simulators
    function of PFC concept.
    Bench-scale validation of PFC concept through submodule testing in        Larger-scale facilities for submodule
4   lab environment simulating heat fluxes or particle fluxes at              testing, High-temperature + all expected
    prototypical levels over long times.                                      range of conditions
    Integrated module testing of the PFC concept in an environment            Integrated large facility:
5   simulating the integration of heat fluxes and particle fluxes at          Prototypical plasma particle flux+heat
    prototypical levels over long times.                                      flux (e.g. an upgraded DIII-D/JET?)
    Integrated testing of the PFC concept subsystem in an environment         Integrated large test facility with
6   simulating the integration of heat fluxes and particle fluxes at          prototypical plasma particle and heat flux
    prototypical levels over long times.
                                                                              Fusion machine
7 Prototypic PFC system demonstration in a fusion machine.                    ITER (w/ prototypic divertor), CTF
    Actual PFC system demonstration & qualification in a fusion reactor       Demo
8 over long operating times.
    Actual PFC system operation to end-of-life in fusion reactor with         Commercial power plant
9 prototypical conditions and all interfacing subsystems.
The level of readiness depends on the design concept
                      Issue-Specific Description                                      Program Elements
    System studies to define tradeoffs and requirements on heat flux level,   Design studies, basic research
1   particle flux level, effects on PFC's (temperature, mass transfer).
    PFC concepts including armor and cooling configuration explored.          Code development, applied research
2   Critical parameters characterized.
     Power plant relevant high-temperature gas-cooled PFC’s
    Data from coupon-scale heat and particle flux experiments; modeling       Small-scale facilities:
3   of governing heat and mass transfer processes as demonstration of         e.g., e-beam and plasma simulators
    function of PFC concept.
    Bench-scale validation of PFC concept through submodule testing in        Larger-scale facilities for submodule
4   lab environment simulating heat fluxes or particle fluxes at              testing, High-temperature + all expected
    prototypical levels over long times.                                      range of conditions
    Integrated module testing of the PFC concept in an environment            Integrated large facility:
5   simulating the integration of heat fluxes and particle fluxes at          Prototypical plasma particle flux+heat
    prototypical levels over long times.                                      flux (e.g. an upgraded DIII-D/JET?)
    Integrated testing of the PFC concept subsystem in an environment         Integrated large test facility with
6   simulating the integration of heat fluxes and particle fluxes at          prototypical plasma particle and heat flux
    prototypical levels over long times.
     Low-temperature water-cooled PFC’s                                       Fusion machine
7 Prototypic PFC system demonstration in a fusion machine.                    ITER (w/ prototypic divertor), CTF
    Actual PFC system demonstration & qualification in a fusion reactor       Demo
8 over long operating times.
    Actual PFC system operation to end-of-life in fusion reactor with         Commercial power plant
9 prototypical conditions and all interfacing subsystems.
    The current status was evaluated for a
        reference ARIES power plant
   For the sake of illustration, we considered a Demo based on the
    ARIES advanced tokamak DCLL power plant design concept
   He-cooled W divertor, DCLL blanket @700˚C, Brayton cycle, plant
    availability=70%, 3-4 FPY in-vessel, waste recycling or clearance
                                                                                 Level completed
                                                                                 Level in progress

                                                               TRL
                                               1   2   3   4    5    6   7   8     9
       Power management
       Plasma power distribution
       Heat and particle flux handling
       High temperature and power conversion
       Power core fabrication
       Power core lifetime
       Safety and environment
       Tritium control and confinement
       Activation product control
       Radioactive waste management
       Reliable/stable plant operations
       Plasma control
       Plant integrated control
       Fuel cycle control
       Maintenance
       The ITER program contributes in
        some areas, very little in others
   ITER promotes to level 6 issues related to plasma and safety
   ITER helps incrementally with some issues, such as blankets,
    PMI, fuel cycle
   The absence of reactor-relevant technologies severely limits its
    contribution in several areas
                                                              TRL
                                              1   2   3   4    5    6   7   8   9
      Power management
      Plasma power distribution
      Heat and particle flux handling
      High temperature and power conversion
      Power core fabrication
      Power core lifetime
      Safety and environment
      Tritium control and confinement
      Activation product control
      Radioactive waste management
      Reliable/stable plant operations
      Plasma control
      Plant integrated control
      Fuel cycle control
      Maintenance
Major gaps remain for several of the key
   issues for practical fusion energy
   A range of nuclear and non-nuclear facilities are required
    to advance from the current status to TRL6
   One or more test facilities such as CTF are required before
    Demo to verify performance in an operating environment

                                                            TRL
                                            1   2   3   4    5    6   7   8   9
    Power management
    Plasma power distribution
    Heat and particle flux handling
    High temperature and power conversion
    Power core fabrication
    Power core lifetime
    Safety and environment
    Tritium control and confinement
    Activation product control
    Radioactive waste management
    Reliable/stable plant operations
    Plasma control
    Plant integrated control
    Fuel cycle control
    Maintenance

								
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