FNS related activities in the ARIES program

FNS related activitiesin 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

Fingers without W/FS joints

Pin-fin cooling

Finger/platecombinations

The T-tube divertor is limited by W temperature more than stress, so tapering is helpful

Tapering reduces temperature and increases stress

Pin-fin experiments demonstrated very high heat removal

• Increases qmax to 18 MW/m2 at expected Re, and to 19 MW/m2 at higher Re

• Allows operation at lower Re for a given qmax lower pressure drop

For plate divertor, pin-fin array

5

qm

ax [

MW

/m2

]

Re (/104)

2 mm Bare 2 mm Pins 0.5 mm Bare 0.5 mm Pins

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

past present

future

ASME code (3Sm)

½ ue

(necking)creep, crack growth,

irradiation effects

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

stress-free-temperature is 1050 ºC

Heat Flux

Elastic Analysis Plastic Analysis

Stress inODS FS

MPa

Stress in F82HMPa

Stress in ODS FS

MPa

Stress inF82HMPa

0.0 MW/m2 (390˚C warm

shutdown)1530 1520 483 370

1.0 MW/m2 1140 1500 231 202

2.0 MW/m2 1520 1510 315 271

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 stableF82H

ODS steel

W

F82H

ODS steel

W

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

ASC VASST

Determination of an optimum design point.

Single-parameter scans around the design point.

Difficult to use, maintain and modify.

Non-interactive tool for self-consistency and costing.

Multi-dimensional parameter space scans.

Large database of physics operating points stored.

Graphical user interface.

Interactive tool for concept exploration.

Fusion Eng. & Design, 85 (2), 243-265, 2010.

Systems analysis flow

Plasmas that satisfy

power and particle balance

Inboard radial build

and engineering

limits

Top and outboard build,

costing

physics engineering build out & costing

Scan several plasma parameters to generate large

database of physics operating points

Screen physics operating points thru

physics filters, engineering

feasibility, and engineering filters

Surviving feasible operating points are built out and costed, graphical display of

parameters (e.g. COE)

1. Toroidal magnetic fields

2. Heat flux to divertor3. Neutron wall load4. Net electric power

Filters include e.g.

VASST GUI v.2

Pull-down menus for common parameters

Blanket database used

Number of points in database

Constraint parameter can restrict database

Auto-labeling

Correlation coefficient

Save plot as TIFF, JPEG, BMP, PNG…

Color bar scale

Edit plotting properties

(Visual ARIES Systems Scanning Tool)

Turn on ARIES-AT point design for reference

The code has multiple applicationsscan 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 N in the aggressive/aggressive

corner

We adopted “technology readiness levels” as the basis for the evaluation

of progress

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

TRL’s express increasing levels of integration and environmental relevance, terms which must be

defined for each technology application

Fusion Science & Tech 56 (2) August 2009.

Utility Advisory Committee“Criteria for practical fusion power

systems”

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

J. Fusion Energy 13 (2/3) 1994.

These criteria for practical fusion suggest three categories of technology readiness

A. Power management for economic fusion energy1. Plasma power distribution2. Heat and particle flux management3. High temperature operation and power conversion4. Power core fabrication5. Power core lifetime

B. Safety and environmental attractiveness6. Tritium control and confinement7. Activation product control and confinement8. Radioactive waste management

C. Reliable and stable plant operations9. Plasma control10.Plant integrated control11.Fuel cycle control12.Maintenance

• LWR spent fuel processing• Waste form development• Fast reactor spent fuel

processing • Fuel fabrication • Fuel performance

cf. GNEP issues:

Example TRL table: Heat & particle flux handling

Issue-Specific Description Program Elements

1System studies to define tradeoffs and requirements on heat flux level, particle flux level, effects on PFC's (temperature, mass transfer).

Design studies, basic research

2PFC concepts including armor and cooling configuration explored. Critical parameters characterized.

Code development, applied research

3Data from coupon-scale heat and particle flux experiments; modeling of governing heat and mass transfer processes as demonstration of function of PFC concept.

Small-scale facilities:e.g., e-beam and plasma simulators

4Bench-scale validation of PFC concept through submodule testing in lab environment simulating heat fluxes or particle fluxes at prototypical levels over long times.

Larger-scale facilities for submodule testing, High-temperature + all expected range of conditions

5Integrated module testing of the PFC concept in an environment simulating the integration of heat fluxes and particle fluxes at prototypical levels over long times.

Integrated large facility:Prototypical plasma particle flux+heat flux (e.g. an upgraded DIII-D/JET?)

6Integrated testing of the PFC concept subsystem in an environment simulating the integration of heat fluxes and particle fluxes at prototypical levels over long times.

Integrated large test facility with prototypical plasma particle and heat flux

7 Prototypic PFC system demonstration in a fusion machine.Fusion machineITER (w/ prototypic divertor), CTF

8Actual PFC system demonstration & qualification in a fusion reactor over long operating times.

Demo

9Actual PFC system operation to end-of-life in fusion reactor with prototypical conditions and all interfacing subsystems.

Commercial power plant

The level of readiness depends on the design concept

Issue-Specific Description Program Elements

1System studies to define tradeoffs and requirements on heat flux level, particle flux level, effects on PFC's (temperature, mass transfer).

Design studies, basic research

2PFC concepts including armor and cooling configuration explored. Critical parameters characterized.

Code development, applied research

3Data from coupon-scale heat and particle flux experiments; modeling of governing heat and mass transfer processes as demonstration of function of PFC concept.

Small-scale facilities:e.g., e-beam and plasma simulators

4Bench-scale validation of PFC concept through submodule testing in lab environment simulating heat fluxes or particle fluxes at prototypical levels over long times.

Larger-scale facilities for submodule testing, High-temperature + all expected range of conditions

5Integrated module testing of the PFC concept in an environment simulating the integration of heat fluxes and particle fluxes at prototypical levels over long times.

Integrated large facility:Prototypical plasma particle flux+heat flux (e.g. an upgraded DIII-D/JET?)

6Integrated testing of the PFC concept subsystem in an environment simulating the integration of heat fluxes and particle fluxes at prototypical levels over long times.

Integrated large test facility with prototypical plasma particle and heat flux

7 Prototypic PFC system demonstration in a fusion machine.Fusion machineITER (w/ prototypic divertor), CTF

8Actual PFC system demonstration & qualification in a fusion reactor over long operating times.

Demo

9Actual PFC system operation to end-of-life in fusion reactor with prototypical conditions and all interfacing subsystems.

Commercial power plant

Power plant relevant high-temperature gas-cooled PFC’s

Low-temperature water-cooled PFC’s

The current status was evaluated for a reference ARIES power plant

  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                  

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

The ITER program contributes in some areas, very little in others

        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                  

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

Major gaps remain for several of the key issues for practical fusion energy

        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                  

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